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Advanced Reactors

2010;():1-10. doi:10.1115/ICONE18-29108.

Establishment of a safety evaluation method is one of the key issues for the nuclear hydrogen production demonstration since fundamental differences in the safety philosophy between nuclear plants and chemical plants exist. In the present study, a practical safety evaluation method, which enables to design, construct and operate hydrogen production plants under conventional chemical plant standards, is proposed. An event identification for the HTTR-IS nuclear hydrogen production system is conducted in order to select abnormal events which would change the scenario and quantitative results of the evaluation items from the existing HTTR safety evaluation. In addition, a safety analysis is performed for the identified events. The results of safety analysis for the indentified five Anticipated Operational Occurrences (AOOs) and three ACciDents (ACDs) show that evaluation items such as a primary cooling system pressure, temperatures of heat transfer tubes at pressure boundary, etc., do not exceed the acceptance criteria during the scenario. In addition, the increase of peak fuel temperature is small in the most severe case, and therefore the reactor core was not damaged and cooled sufficiently. These results will contribute to the safety review from the government and demonstration of the nuclear production of hydrogen.

Commentary by Dr. Valentin Fuster
2010;():11-12. doi:10.1115/ICONE18-29206.

Current HTGRs such as the High Temperature Engineering Test Reactor (HTTR) of Japan Atomic Energy Agency (JAEA) use Tri-Isotropic (TRISO)-coated fuel particles with diameter of around 1 mm. TRISO fuel consists of a micro spherical kernel of oxide or oxycarbide fuel and coating layers of porous pyrolytic carbon (buffer), inner dense pyrolytic carbon (IPyC), silicon carbide (SiC) and outer dense pyrolytic carbon (OPyC). The principal function of these coating layers is to retain fission products within the particle. Particularly, the SiC coating layer acts as a barrier against the diffusive release of metallic fission products and provides mechanical strength for the particle [1].

Commentary by Dr. Valentin Fuster
2010;():13-17. doi:10.1115/ICONE18-29214.

A sodium-cooled MOX-fueled FBR core concept to improve nuclear proliferation resistance was proposed. First, we set an index for the nuclear proliferation resistance. In a previous study, reactor-grade Pu was defined such that the Pu-240 isotopic ratio was larger than 18%. Another study defined nuclear proliferation resistance with the Pu-238 isotopic ratio considering its higher spontaneous fission rate and decay heat. We tentatively use the total isotope composition ratio of Pu-238 and Pu-240 as a proliferation resistance index in line with the earlier studies. Next, we designed the sodium-cooled mixed-oxide (MOX)-fueled core concept with the breeding ratio (BR) of over 1.1 without a radial blanket. To attain the index for nuclear proliferation resistance, we added minor actinides (MAs) to the axial blanket fuel (AB). Contents of MAs in the AB to achieve the proliferation resistance index were evaluated. For the case of Np as a representative MA, the minimum content of Np to achieve the index was 3%. And, for the case of loading all MAs, the minimum content of MAs was 10.5%.

Commentary by Dr. Valentin Fuster
2010;():19-25. doi:10.1115/ICONE18-29231.

China Advanced Research Reactor (CARR), which will be critical in China Institute of Atomic Energy (CIAE) in 2010, is a multipurpose, high neutron flux and tank-type (inverse neutron trap) reactor with compact core. Its nominal reactor power is 60MW and the maximum thermal neutron flux is about 8.0×1014 n/cm2 ·s in heavy water tank. It has a cylindrical core having a diameter of about 450mm and a height of 850mm. The CARR’s core consists of seventeen plate-type standard fuel elements and four follower fuel elements, initially loaded with 10.97 kg of 235 U. The fuel element has been designed with U3 S2 -Al dispersion containing 235 U of (19.75±0.20)wt.% low enriched uranium (LEU) and having a density of 4.3gU/cm3 . The aluminum alloy is used as the cladding. There are twenty-one and seventeen fuel plates in the standard and follower fuel element, respectively. There are specific requirements for design of the fuel element and strict limitation for the operation parameters due to the high heat flux and high velocity of coolant in CARR. Irradiation test of fuel element had been carried out at fuel element power of 3.1±20%MW at Russia MIR reactor. Average burnup of fuel element is up to 40%. This paper deals with the detailed design of fuel element for CARR, out-pile and in-pile test projects, including selection of fuel and structure material, description of element structure, miniplates and fuel element irradiation experiment, measurement of properties of fuel plate, fabrication of fuel element and test results.

Topics: Fuels , Design , China
Commentary by Dr. Valentin Fuster
2010;():27-33. doi:10.1115/ICONE18-29248.

This paper resumes the results of the collaboration between AREVA and CNPDC during the past two years for performing and achieving the basic design of EPR™ reactor CVCS system for the TSN NPP. The CVCS (Chemical and Volume Control System) is an essential auxiliary system of the PWR technology based nuclear power plants all around the world. In the EPR™ reactor design, as it is also in similar nuclear power plants, this auxiliary system has well determined functions, which are: reactivity control, reactor coolant volume control, coolant chemistry control, primary system main pumps seal water injection as well as the pressurizer auxiliary spray regulation for the Reactor Coolant System. In the EPR™ reactor design, the CVCS is mainly an operational system and only some valves and instrumentations take part at some specific safety functions, (e.g. Reactivity Control, Containment of Radioactive Substances). In the first part of this paper a general introduction to the EPR™ reactor CVCS technology, including the related safety functions and detailed operational functions of CVCS, is presented. In the TSN EPR™ reactor CVCS design, the system is divided into eight sections, (defined from RCV1 to RCV8). The corresponding detailed description of these sections, including their functions, structure and main components, as they have been implemented in the EPR™ reactor CVCS design for the TSN NPP, is then presented in the second part of this paper. In addition some specific design features for EPR™ reactor CVCS system for the TSN NPP, such as the hydrogenation station technology, are also focused in this paper. The reference power plant, concerning the CVCS design, for the TSN NPP is the FA3 NPP, but different design concepts have been implemented in the TSN NPP with regards to the coolant purification section (RCV2), and the coolant filtering in the reactor coolant pumps seal injection and leak-off lines.

Topics: Design
Commentary by Dr. Valentin Fuster
2010;():35-42. doi:10.1115/ICONE18-29290.

Sorption of fission product vapors on metallic surfaces and dust particles is an important safety aspect of HTR reactors. Safety analyses of these reactors are performed using computer codes, such as MELCOR, RADAX, SPECTRA. These codes have sorption models allowing to compute the sorption rates of different fission products on surfaces. The code users must supply the model coefficients applicable for the particular surface and isotope. This paper describes the work performed to find relevant experimental data and find the sorption coefficients that represent well the available data for iodine on different surfaces. The purpose of this work is to generate a set of coefficients that may be recommended for the computer code users. Calculations were performed using the computer code SPECTRA. The following data was analyzed: • Sorption of I-131 on graphite; • Sorption of I-131 on steel; • Sorption of I-131 on dust. The results are summarized as follows: • The available data is provided in form of Langmuir isotherms. • The Langmuir isotherms do not provide sufficient data to define all sorption coefficients. The Langmuir isotherm provides equilibrium data; the relaxation time (to get to equilibrium) needs to be guessed. In practice this means that one of the sorption coefficients must be guessed. In the present calculations the desorption coefficient was being guessed and then varied in sensitivity calculations. The calculations showed that surface concentration is not sensitive to the choice of the parameter. • The sorption model in SPECTRA is capable to correctly reproduce the sorption behavior given by the Langmuir isotherms. • Out of the calculated cases, the highest activity (surface concentration) is observed on the steel surface; the lowest on the graphite surfaces. • The present work may serve as a useful guide of how to convert the Langmuir isotherm data into the input parameters required for computer code calculations.

Topics: Sorption
Commentary by Dr. Valentin Fuster
2010;():43-52. doi:10.1115/ICONE18-29371.

Supercritical carbon dioxide (S-CO2 ) gas turbines can generate power at high cycle thermal efficiency, even at modest temperatures of 500–550°C, because of their markedly reduced compressor work near the critical point. Furthermore, the reaction between Na and CO2 is milder than that between H2 O and Na. A more reliable and economically advantageous power generation system could be achieved by coupling with a sodium-cooled fast reactor. At Tokyo Institute of Technology, numerous development projects have been conducted for development of this system in cooperation with JAEA. Supercritical CO2 compressor performance test results are given as described herein. A centrifugal compressor is chosen for the performance test. Main compressor parts are stored in a pressure vessel. Maximum design conditions of the supercritical CO2 test apparatus are pressure of 11 MPa, temperature of 150°C, the flow rate of 6 kg/s and rotational speed of 24,000 rpm. The centrifugal compressor has an electric motor with permanent magnets on the rotor surface, with speed control by an inverter up to 24,000 rpm, a rotor shaft for the impeller, and a motor supported by gas bearings. Different compressor design points are examined using impellers of three kinds; test data are obtained using those impellers under steady state conditions with changing pressure, temperature, flow rate, and compressor rotor speed. The pressure ratio (compressor outlet pressure/inlet pressure) is obtained with the function of compressor rotational speed and the fluid flow rate. The data cover a broad region from sub-critical to supercritical pressure. Such data were obtained for the first time. No unstable phenomenon was observed in the area where the CO2 properties change sharply. Data of the pressure ratio vs. flow rate were coincident with the fundamental compressor theory.

Commentary by Dr. Valentin Fuster
2010;():53-58. doi:10.1115/ICONE18-29379.

The high temperature isolation valve (HTIV) is a key component to assure the safety of a high temperature gas cooled reactor (HTGR) connected with a hydrogen production system, that is, protection of radioactive material release from the reactor to the hydrogen production system and combustible gas ingress to the reactor at the accident of fracture of an intermediate heat exchanger and the chemical reactor. The HTIV used in the helium condition over 900 °C, however, has not been made for practical use yet. The conceptual structure design of an angle type HTIV was carried out. A seat made of Hasteloy-XR is welded inside a valve box. Internal thermal insulation is employed around the seat and a liner because high temperature helium gas over 900 °C flows inside the valve. Inner diameter of the top of seat was set 445 mm based on fabrication experiences of valve makers. A draft overall structure was proposed based on the diameter of seat. The numerical analysis was carried out to estimate temperature distribution and stress of metallic components by using a three-dimensional finite element method code. Numerical results showed that the temperature of the seat was simply decreased from the top around 900 °C to the root, and the thermal stress locally increased at the root of the seat which was connected with the valve box. The stress was lowered below the allowable limit 120 MPa by decreasing thickness of the connecting part and increasing the temperature of valve box to around 350 °C. The stress also increased at the top of the seat. Creep analysis was also carried out to estimate a creep-fatigue damage based on the temperature history of the normal operation and the depressurization accident.

Commentary by Dr. Valentin Fuster
2010;():59-63. doi:10.1115/ICONE18-29392.

Primary pipe rupture is an important design base accident in the high temperature gas cooled reactor (HTGR). When a primary pipe of the HTGR ruptures, helium coolant gas in the reactor blows out into the reactor confinement structure and the reactor primary system depressurizes. After the pressures of the reactor and the confinement equalizes, air is expected to enter the reactor core from the breach. Consequently, the core graphite structures may be oxidized by the air and the complicated natural convection of multi component gas mixtures with chemical reactions would take place inside the reactor. Hence, it is necessary to investigate the air ingress process, the natural convection of multi component gas mixtures in order to understand the effect and develop mitigation of the air ingress. JAEA has performed analysis and fundamental experiments about air ingress from the rupture of one or more main coolant pipes on the lower body of the RPV. These studies showed the air ingress phenomena in the depressurized reactor and proposed a new passive mechanism of sustained counter air diffusion (SCAD) that has been shown effective in preventing major air ingress through natural circulation in the reactor. In the present plan, JAEA will construct an experimental reactor mockup including reactor core, the SCAD system, pressure vessel, coaxial pipe and so on. The core is made of graphite or ceramics and heated by electric heaters to allow for test operation up to 1200°C. Present status of these activities will be presented. Based on the analytical results and know-how obtained through the bench test, a 1/8 scale air ingress mockup test, which intends to simulate the accident condition of GTHTR300, is being planned with a conceptual design as the next step of the air ingress experiment evaluation in JAEA. In the design of mockup experimental facility, it is important to reproduce flow phenomena in a reasonable scale from the viewpoint of construction cost. We designed the internal structure to reproduce mixing performance of multi-component flow involving ingress phenomena especially in the guillotine breaks of primary coolant pipe. Complex flow pattern with gas oxidized chemical interaction in the graphite porous structure of the HTGR core will be characterized. Preliminary analytical results especially natural circulation flow patterns induced by density and concentration difference obtained with a CFD model agreed well with that measured by bench experiments, which showed natural circulation pattern in a simplified reactor.

Commentary by Dr. Valentin Fuster
2010;():65-77. doi:10.1115/ICONE18-29399.

Due to global warming and high oil prices nuclear power is the most feasible solution for generating electricity. For the fledging nuclear power industry small and medium sized nuclear reactors (SMR’s) are instrumental for the development and demonstration of nuclear reactor technology. Due to the enhanced and outstanding safety features, these reactors have been considered globally. In this paper, first we have summarized the reactor design by considering some of the large nuclear reactor including advanced and theoretical nuclear reactor. Secondly, comparison between large nuclear reactors and SMR’s have been discussed under the criteria led by International Atomic Energy Agency (IAEA). Thirdly, a brief review about the design and safety aspects of some of SMR’s have been carried out. We have considered the specifications and parametric analysis of the reactors like: ABV which is the floating type integral Pressurized water reactor; Long life, Safe, Simple Small Portable Proliferation Resistance Reactor (LSPR) concept; Multi-Application Small Light Water Reactor (MASLWR) concept; Fixed Bed Nuclear Reactor (FBNR); Marine Reactor (MR-X) & Deep Sea Reactor (DR-X); Space Reactor (SP-100); Passive Safe Small Reactor for Distributed energy supply system (PSRD); System integrated Modular Advanced Reactor (SMART); Super, Safe, Small and Simple Reactor (4S); International Reactor Innovative and Secure (IRIS); Nu-Scale Reactor; Next generation nuclear power plant (NGNP); Small, Secure Transportable Autonomous Reactor (SSTAR); Power Reactor Inherently Safe Module (PRISM) and Hyperion Reactor concept. Finally we have point out some challenges that must be resolved in order to play an effective role in Nuclear industry.

Commentary by Dr. Valentin Fuster
2010;():79-84. doi:10.1115/ICONE18-29412.

The High Temperature Engineering Test Reactor (HTTR) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR is a graphite-moderated and helium gas-cooled reactor with thermal power of 30MW and the maximum reactor outlet coolant temperature of 950°C. Main objectives of the HTTR are to establish and develop HTGR technology and to demonstrate process heat application. The HTTR has conducted two test operations which are safety demonstration test and continuous operation. The safety demonstration tests focus on the verification of the inherent safety features of the HTGR that is the negative reactivity feedback effect of the core brings the reactor power safely to a safe and stable level without a reactor scram and the temperature transient of the reactor is slow in case of anticipated operational occurrences (AOOs). The safety demonstration tests include reactivity insertion test, coolant flow reduction test and loss of forced cooling (LOFC) test. Reactivity insertion test and coolant flow reduction test have been conducted since 2002. These tests demonstrated the inherent safety features of the HTGR in case of reactivity insertion and coolant flow reduction, and provided valuable data for code validation. LOFC test will start in the middle of 2010. LOFC is one of the important accident scenarios in the safety assessment of the HTGR. This test result will show extreme safety features of the HTGR and further improve the safety design approach of the HTGR. Obtained data can be useful to validate plant safety analysis codes. The continuous operation has been conducted to obtain plant data to establish HTGR technology and to demonstrate capability of the HTTR to supply stable heat to heat utilization system for long-term. Two operations of 30-day continuous operation in rated operation mode (in which designed reactor outlet coolant temperature of 850°C) and 50-days continuous operation in high temperature test operation mode (in which designed reactor outlet coolant temperature of 950°C) have been conducted so far. The 30-day continuous operation was achieved in 2007 and a good fuel performance to retain fission products within the coated fuel particle was clarified. The HTTR conducts 50-days continuous operation in 2010 to add useful operation data at high temperature to improve technical basis of HTGR and to realize high temperature heat application of HTGR.

Commentary by Dr. Valentin Fuster
2010;():85-89. doi:10.1115/ICONE18-29521.

In the entire history of commercial nuclear power so far, only two major accidents leading to damage of reactor core have taken place. One is Three Mile Island (TMT) accident (1979), which is caused by a series of human error, and the other is Chernobyl accident (1986), which is due to the combined reason of design defects and human errors. After TMI and Chernobyl accidents, in order to reduce manpower in operation and maintenance and influence of human errors on reactor safety, consideration is given to utilization of passive safety systems. According to the IAEA definition, passive safety systems are based on natural forces, such as convection and gravity, and stored energy, making safety functions less dependent on active systems and operators’ action. Recently, the technology of passive safety has been adopted in many reactor designs, such as AP1000, developed by Westinghouse and EP1000 developed by European vendor, and so on. AP1000 as the first so-called Generation III+ has received the final design approval from US NRC in September 2004, and now being under construction in Sanmen, China. In this paper, the major passive safety systems of AP1000, including passive safety injection system, automatic depressurization system passive residual heat removal system and passive containment cooling system, are described and their responses to a break loss-of-coolant accident (LOCA) are given. Just due to these passive systems’ adoption, the nuclear plant can be able to require no operator action and offsite or onsite AC power sources for at least 72h when one accident occurs, and the core melt and large release frequencies are significantly below the requirement of operating plants and the NRC safety goals.

Commentary by Dr. Valentin Fuster
2010;():91-96. doi:10.1115/ICONE18-29537.

Physical characteristics of the coolant in the Supercritical-pressure Light Water Cooled Reactor (SCWR) vary greatly near the pseudo-critical point, which will cause variations of core neutron cross section and then bring about power perturbation. Further it will prompt the corresponding thermal parameters of supercritical water changed, and form feedback action, finally resulting in intensely coupled thermal-hydraulics and neutron-physical. Proper fuel assembly has been selected as research object, and the model of multiple parallel channels has been established. In view of this model, using DRAGON code for neutron-physical calculations and developing corresponding thermal-hydraulic programs, and then achieve coupling them through appropriate data interface, the calculation platform established. Finally the power distribution and the corresponding parameters temperature distributions in the model have been predicted. On account of deficiencies reflected in calculations, such as the heterogeneous power distribution, fuel assembly geometry has been changed, for instance the proper peripheral moderator wall has been added based on the preceding assembly, then do the coupling calculations and analyze the results. Comparisons between different results have been made, and the expected aim has been reached, which can prove the rationality of assembly modifications and meanwhile prove the usability of the calculation platform. Thus, modified assembly and the calculation platform could be the calculation foundation in future designs of SCWR.

Commentary by Dr. Valentin Fuster
2010;():97-101. doi:10.1115/ICONE18-29542.

Heat transfer coefficient is an important feature factor in describing SCWR. Using grey theory and multiple regression, writing code with MATLAB with considering temperature and enthalpy of coolant, heat flux and system pressure, establish GM(1.1), GM(1,3) and multiple regression models of heat transfer coefficient of supercritical water. analyzing the impact of temperature and enthalpy of coolant, heat flux, pressure on the changes of heat transfer coefficient in SCWR. Grey model generally summarize the experimental data. and its calculation results are compared with the results of regression model, which shows grey model can forecast the changes of heat transfer coefficient better, provides new methods of fitting and forecasting heat transfer coefficient of supercritical water.

Commentary by Dr. Valentin Fuster
2010;():103-114. doi:10.1115/ICONE18-29658.

Safety analyses for the XT-ADS were performed with the reactor safety code SIMMER-III. Besides a brief description of the numerical model, three typical transients are presented in this paper, namely, the unprotected loss of flow (ULOF), unprotected transient over-current (UTOC), and the unprotected coolant flow blockage accident (UBA). Because of the important phenomenon of mass flow rate undershooting in the ULOF case, an integral equation model was set up for a further theoretical study of ULOF. The model confirms the numerical simulation results for various cases and gives a deeper understanding of this phenomenon. The faster the pump shut down, the larger is the undershooting of the mass flow rate. On the other hand a larger coolant cold leg area leads to a weaker undershooting. The stability analysis shows that the natural convection state is in the region of the damped oscillation for the current XT-ADS design.

Commentary by Dr. Valentin Fuster
2010;():115-118. doi:10.1115/ICONE18-29660.

According to the similarity law, the experimental system of pebble dynamics of two dimensions reactor core in HTR can be established. With the execution of phenomenological research of the moving rules of pebbles, it is available to obtain distribution of path lines, curves of degree of dispersion, and distribution of stagnant time. The results turn out to be that (1) the motion of a single pebble is partially diffused and random in micro view, but the rule of all pebbles is assured and steady in macro view; (2) curves of degree of dispersion continue to increases at first and then decrease gradually. The degree of dispersion of bilateral pebbles is bigger than that of central ones; (3) stagnant time is mainly linear to height and the ratio of stagnant time of radical location and central location increase when a pebble is moving down.

Commentary by Dr. Valentin Fuster
2010;():119-122. doi:10.1115/ICONE18-29664.

The CANDU® reactor is the most resource-efficient reactor commercially available. The features that enable the CANDU reactor to utilize natural uranium facilitate the use of a wide variety of thorium fuel cycles. In the short term, the initial fissile material would be provided in a “mixed bundle”, in which low-enriched uranium (LEU) would comprise the outer two rings of a CANFLEX® bundle, with ThO2 in the central 8 elements. This cycle is economical, both in terms of fuel utilization and fuel cycle costs. The medium term strategy would be defined by the availability of plutonium and recovered uranium from reprocessed used LWR fuel. The plutonium could be used in Pu/Th bundles in the CANDU reactor, further increasing the energy derived from the thorium. Recovered uranium could also be effectively utilized in CANDU reactors. In the long term, the full energy potential from thorium could be realized through the recycle of the U-233 (and thorium) in the used CANDU fuel. Plutonium would only be required to top up the fissile content to achieve the desired burnup. Further improvements to the CANDU neutron economy could make possible a very close approach to the Self-Sufficient Equilibrium Thorium (SSET) cycle with a conversion ratio of unity, which would be completely self-sufficient in fissile material (recycled U-233).

Commentary by Dr. Valentin Fuster
2010;():123-130. doi:10.1115/ICONE18-29683.

Graphite blocks are important core components of the high temperature gas-cooled reactor. As these blocks are simply stacked in array, collisions among neighboring components may occur during earthquakes or accidents. Thus, it is important to develop a reliable seismic model of the stacked graphite blocks and have them designed to maintain their structural integrity during the anticipated occurrences. Various aspects involved in modeling and calculating impact-contact dynamics can affect the resulting behavior of the graphite block. These include mesh size, time step, contact behavior, mechanical constraint formulation of impact-contact analysis, etc. This work is dedicated to perform comparative studies and the effects of these parameters will be identified. The insights obtained through these studies will help build a realistic impact-contact model of the graphite block, from which a lumped or reduced dynamics model will be developed for the seismic analysis of the reactor including these graphite components.

Commentary by Dr. Valentin Fuster
2010;():131-138. doi:10.1115/ICONE18-29694.

The Lead Cooled Fast Reactor (LFR) is one of the six concepts selected by the Generation IV International Forum (GIF) as candidates for the long term evolution of nuclear technology. Due to the significant technological innovations it implies, the European Sustainable Nuclear Energy Technology Platform (SNETP) recognized that LFR complete development requires the realization of a demonstration plant (DEMO) as a fundamental intermediate step. In this paper, a preliminary approach to the simulation of DEMO primary system dynamic behavior is presented. The need of investigating reactor responses to temperature transients has led to a simplified model reckoning with all the main feedbacks following a reactivity change in the core, which have been calculated by means of ERANOS deterministic code ver. 2.1 coupled with JEFF3.1 data library. A lumped-parameter approach has been adopted to treat both neutronics and thermal-hydraulics: indeed, the point-kinetics approximation has been employed and an average-temperature heat-exchange model has been implemented. Due to the latter, the dynamic mechanical behavior of DEMO core — modeled as a cylinder — has been addressed by considering expansions and contractions instantaneous with temperature variations, i.e. neglecting mass inertia effects. The very simple linearized model treated in the present work turns out to be a helpful tool in this early phase of the DEMO pre-design, in which all the system specifications are still considered to be open design parameters, since it allows a relatively quick, qualitative analysis of dynamics and stability aspects that cannot be left aside when refining or even finalizing the system configuration.

Commentary by Dr. Valentin Fuster
2010;():139-143. doi:10.1115/ICONE18-29744.

One of the most important components of a linac is buncher. Throughout most of the buncher, the electrons are well forward of the crest and have velocities considerably less than light velocity, thus they are in a region of radial defocusing and a considerable fraction of the beam will be lost unless defocusing action is counteracted by some other applied forces. The simplest way to do this is to set up a longitudinal magnetic field which interacts with the radial motion of the electrons and causes them to follow helical orbits through the space occupied by the field. In this paper, five solenoids were designed to provide necessary magnetic field inside the buncher. Magnetic field was analytically calculated and compared with simulation results of CST. Because of resistance in wires, some amount of energy appears in form of heat, so heat power was calculated analytically and cooling system was designed for these solenoids by ANSYS.

Commentary by Dr. Valentin Fuster
2010;():145-154. doi:10.1115/ICONE18-29755.

This paper presents current status and development of nuclear calculation methodology for sodium cooled fast reactor in Mitsubishi and two core design codes; an original two dimensional cell/lattice calculation code, PIJHEX and a thermal hydraulic calculation code, MIX-MKII. Mitsubishi Atomic Power Industries, Inc. developed several core design codes for Monju and JOYO and recently this activity continues still further for the next FBR in Japan. It is explained in this paper that the PIJHEX is confirmed to be valid from the comparison of Monte Carlo code and the examination on physical phenomenon in calculation results. Besides, the approach toward design and development of code system in Mitsubishi and MIX-MKII are introduced briefly.

Commentary by Dr. Valentin Fuster
2010;():155-163. doi:10.1115/ICONE18-29780.

CANDU-SCWR is one Generation IV reactor being developed in Canada. Significant amount of efforts has been made to develop CANDU-SCWR. Little work has been done on the dynamic analysis and control design. To observe the dynamic behaviours of CANDU-SCWR, the detailed CANDU-SCWR thermal-hydraulic model is developed. The movable boundary method is adopted for CANDU-SCWR thermal-hydraulic modeling. The benefits of adopting movable boundary are derived from the comparisons with the fixed boundary method. The steady-state results agree well with the design data. The responses of CANDU-SCWR reactor to different disturbances are simulated and analyzed and the results are reasonable in theory. Linear dynamic models are derived from simulation data of CANDU-SCWR thermal-hydraulic model around the design operating point using a system identification technique to facilitate the control system design. The linear dynamic models are validated and it is shown that they can describe the dynamic characteristics of CANDU-SCWR around the design point accurately.

Commentary by Dr. Valentin Fuster
2010;():165-170. doi:10.1115/ICONE18-29783.

Nuclear power is a proven, safe, plentiful and clean source of power generation, and AP1000 is the only Generation III+ reactor to receive Design Certification from the U.S. Nuclear Regulatory Commission (NRC). One of the major features of this advanced reactor is the utilization of the passive safety systems and a steel Containment Vessel (CV). As the boundary isolated to keep the radioactive waste to release to outside and also the medium to transfer the heat, CV is a mainly part of the Passive Containment Cooling (PCS) System, which is the key to move the resident heat from the core to the air. The CV is made up with Sub-Assemblies, such as Bottom Head (CVBH), 1st Ring, 2nd Ring, 3rd Ring, and Top Head (CVTH), which are assembled with steel plates. So to build the Containment Vessel on site, we have to assemble each individual part with steel plates fabricated in shop first, transport the Sub-Assembly to Nuclear Island (NI) and lift them in place. Right now, the first AP1000 in the world is under construction at Sanmen, China. Based on the statement above, the construction of CVBH of the 1st AP1000 is a notable achievement and implementation without experience proven, the constructability analysis about work sequence, detailed logic, activity duration, production rate and improvement of them during the process is worthy to study.

Commentary by Dr. Valentin Fuster
2010;():171-179. doi:10.1115/ICONE18-29785.

In this paper, computational fluid dynamics (CFD) gas flow simulations are carried out. In CFD calculations, geometry modeling and physical modeling are crucial to CFD results. The effects of the treatments of the inter-pebble contacts on gas flow fields and heat transfer are examined; a sensitivity analysis for the gap size is conducted with two spherical pebbles, in which the inter-pebble region is modeled by means of two types of inter-pebble gaps and two kinds of direct contacts. On the other hand, both of large eddy simulation (LES) and Reynolds-averaged Navier-Stokes (RANS) models are employed to investigate the turbulent effects. It is found that the flow fields and relevant heat transfer are significantly dependent on the modeling of the inter-pebble regions. The calculations indicate the complex flow structures present within the voids between the fuel pebbles.

Commentary by Dr. Valentin Fuster
2010;():181-186. doi:10.1115/ICONE18-29792.

The construction of AP1000 Nuclear Power Plant (NPP) has commenced in China. The AP1000 NPP features a passive design concept and modular construction technology. Based on the management of the construction of current AP1000 NNP, this paper describes the effects on Nuclear Island (NI) construction project management resulting from modular construction technology, as well as new construction techniques and methods. This paper puts forward new requirements for construction schedule management of the nuclear island construction at different levels. The AP1000 NI construction logic features the parallel construction of civil and structural erection as the main approach, with the integrated schedule of module fabrication, assembly and installation as support.

Commentary by Dr. Valentin Fuster
2010;():187-199. doi:10.1115/ICONE18-29800.

The U.S Department of Energy’s (DOE) Next Generation Nuclear Plant (NGNP) has three goals in its mission: deliver (1) electricity, (2) process steam and heat, and (3) generate hydrogen to support a hydrogen economy. DOE selected the high-temperature, helium-cooled graphite (moderated) reactor (HTGR) technology for its fundamental design. However, the NGNP faces many challenging design, licensing and cost hurdles. With huge risks and its associated expenses, the NGNP — a completely new reactor technology — requires an open-minded understanding, consideration and application of earlier HTGR design features. This paper suggests reconsidering some HTGR features in an updated Generation IV HTGR as a design alternative.

Commentary by Dr. Valentin Fuster
2010;():201-209. doi:10.1115/ICONE18-29807.

Air ingress into a very high temperature gas-cooled reactor (VHTR) is an important phenomenon to consider because the air oxidizes the reactor core and lower plenum where the graphite structure supports the core region in the gas turbine modular helium reactor (GT-MHR) design, thus jeopardizing the reactor’s safety. Validating the computational fluid dynamics (CFD) code used to analyze the air ingress phenomena is therefore an essential part of the safety analysis and the ultimate computation required for licensing. An experimental data set collected by ETH Zurich on a lock exchange experiment (Grobelbauser et al., Lowe et al. 2002; Lowe et al. 2005; and Shin et al. 2004) was selected for the validation. The experiment was based on a series of lock exchange flows with gases of different density ratios varying from 0.046 to 0.9 in a closed channel of a square cross-section. The focus was on the quantitative measurement of front velocities of the gravity current flows. The experiment results cover the full range of gas intrusions—heavy as well as light—for the gravity current flows in the lock exchange situations. FLUENT CFD code (ANSYS Fluent 2008) was used. The calculated results showed very good agreement with the experimental data. A number of tables and comparison plots are included to summarize the estimated current speeds. The current speed obtained by experimental data was 1.25 m/s and that of the simulation was 1.19 m/s. This result indicates that the deviation of the simulation is only 4.8% that of the experimental data.

Commentary by Dr. Valentin Fuster
2010;():211-218. doi:10.1115/ICONE18-29839.

An extensive thermalhydraulics and safety program has been established in Canada to support the research and development effort for the CANDU supercritical water-cooled reactor design. This program focuses mainly on establishing the research capability at various Canadian universities, and includes several collaborative projects with Chinese universities to expedite the pre-conceptual design. Several test facilities in support of the supercritical heat transfer research are described. Selected results are presented from analytical studies.

Commentary by Dr. Valentin Fuster
2010;():219-225. doi:10.1115/ICONE18-29844.

This paper mainly talks about supercritical water cooled reactor (SCWR) fuel assembly which has special character such as moderator channels, two flue pin rows between the moderator channels and the thermal zone of fuel bundle. SABER (Sub-channel Analysis Based on Extended pressure Range) is used for sub-channel analysis of the fuel assembly. It is a self-developed sub-channel thermal-hydraulic analysis code. The code was originally developed to analyze SCWRs. After adding an extended computational cell structure and some new boundary conditions, SABER can simulate the complex flow (the opposite flow direction between moderator channels and coolant channels) in the fuel assembly. SABER can also compute heat exchange between moderator channels and cooling channels. The phenomenon of inverse flow occurs in the moderator channels, was found by sub-channel analysis. The phenomenon is caused by the different heating situation around moderator channels. Some sensitivity analysis of this phenomenon was analyzed in the paper.

Commentary by Dr. Valentin Fuster
2010;():227-232. doi:10.1115/ICONE18-29878.

We provide an update on recent advances in pressure tube reactor (PTR) technology as exemplified by progress in the continuous research and development path that has been defined. This includes numerous and fruitful strategic development partnerships, both nationally and internationally, to contribute to the major areas of technological innovation, economic competitiveness, energy security, sustainability and environmental stewardship. We discuss the major areas of advancement in R&D that utilize the inherent features and advantages of the unique PTR concepts of utilizing a low pressure, heavy water moderator with a modular channel construction.

Topics: Pressure
Commentary by Dr. Valentin Fuster
2010;():233-239. doi:10.1115/ICONE18-29899.

International efforts on materials selection and development related to supercritical water-cooled reactors (SCWRs) have produced a considerable amount of data in the open literature [1], The majority of these data are on aspects of materials properties such as corrosion, stress corrosion cracking, creep, irradiation damage as well as microstructural degradation under various exposure conditions. These prior efforts are helping guide the current selection of candidate alloys for further, longer-term evaluation. As continuing research on the SCWR advances, gaps and limitations in the published data are being identified. In terms of corrosion properties, these gaps can be seen in several areas, including: 1) the test environment, 2) the physical and chemical severity of the tests conducted as compared with likely reactor service/operating condition, and 3) test methods used. While some of these gaps can be filled readily by the current research projects, in particular those occurring in Generation IV International Forum (GIF) member countries, others require further advances in our understanding of material-environment interactions involving supercritical water. Gaps in advanced test facilities for future research are also becoming evident. Future needs for materials development and suggestions for expanded international collaborations to link with groups working on materials for advanced fossil-fired supercritical water power plants, as well as other GEN IV and fusion reactor designs are summarized.

Commentary by Dr. Valentin Fuster
2010;():241-249. doi:10.1115/ICONE18-29910.

The European Facility for Industrial Transmutation (EFIT) is developed to transmute long-lived actinides from spent fuel on an industrial scale. In this lead-cooled reactor an intermediate loop is eliminated for economic reasons. Within the framework of design and safety studies the impact of a steam generator tube rupture accident has been investigated. In this postulated event high-pressured liquid water blasts into the lead pool which could trigger various transients. As a major concern steam could be dragged into the core featuring a positive void worth. A thermal lead/water interaction could lead to in-core damage propagation; it could initiate a sloshing of the lead coolant and trigger voiding processes. Furthermore the pressurization of the cover gas needs to be considered. To prove the feasibility of the proposed design these risks are investigated and assessed. Numerical simulations are performed using the advanced safety analysis code SIMMER-III [2]. For the important issue of thermal lead/water interactions the SIMMER code has been validated against Japanese heavy-liquid/water injection experiments.

Commentary by Dr. Valentin Fuster
2010;():251-258. doi:10.1115/ICONE18-29967.

Japanese national project of next generation light water reactor (LWR) development started in 2008. As one of its development items, the thermal-hydraulic test of spectral shift rod (SSR) is planned. A new component called SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR’s merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies core void fraction change, resulting in the amplified spectral shift effect. In this paper, its test plan overview and pre-test analysis by TRACG code is presented. The test plan was developed with the purpose of evaluating SSR thermal-hydraulic characteristics at the actual BWR operating condition (7MPa), such as the controllability of SSR water level, and obtaining data for the validation of calculation method. In the test plan, several types of SSR simulation which covers SSR design in both next generation BWR and conventional BWR were designed. Also test operating conditions such as thermal-hydraulic parameters are determined. In order to evaluate these test specifications, pre-test analysis by TRACG code was conducted. Analysis results of each parameter’s effect on SSR characteristics are consistent with SSR mechanism, which shows that the actual operating condition for SSR fuel is simulated well.

Commentary by Dr. Valentin Fuster
2010;():259-267. doi:10.1115/ICONE18-29986.

In order to provide a consistent direction to long-term R&D activities, the Korea Atomic Energy Commission (KAEC) approved a long-term development plan for future nuclear reactor systems which include sodium cooled fast reactor (SFR) and very high temperature reactor (VHTR) on December 22, 2008. The SFR system is regarded as a promising technology to perform actinide management. The final goal of the long-term SFR development plan is the construction of an advanced SFR demonstration plant by 2028. The nuclear hydrogen project in Korea aims at designing and constructing a nuclear hydrogen demonstration system by 2022 to demonstrate its hydrogen production capability. This paper summarizes the overall long-term project plans for SFR and VHTR technology development and explains results of detailed design studies with supporting R&D activities.

Commentary by Dr. Valentin Fuster
2010;():269-277. doi:10.1115/ICONE18-29993.

Currently, there are a number of Generation IV SuperCritical Water-cooled nuclear Reactor (SCWR) concepts under development worldwide. The main objectives for developing and utilizing SCWRs are: 1) Increase gross thermal efficiency of current Nuclear Power Plants (NPPs) from 30–35% to approximately 45–50%, and 2) Decrease capital and operational costs and, in doing so, decrease electrical-energy costs. SuperCritical Water (SCW) NPPs will have much higher operating parameters compared to current NPPs (i.e., steam pressures of about 25 MPa and steam outlet temperatures up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. Furthermore, SCWRs operating at higher temperatures can facilitate an economical co-generation of hydrogen through thermo-chemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis. To decrease significantly the development costs of an SCW NPP, to increase its reliability, and to achieve similar high thermal efficiencies as the advanced fossil-fired steam cycles, it should be determined whether SCW NPPs can be designed with a steam-cycle arrangement that closely matches that of mature SuperCritical (SC) fossil-fired thermal power plants (including their SC-turbine technology). The state-of-the-art SC-steam cycles at fossil-fired power plants are designed with a single-steam reheat and regenerative feedwater heating. Due to this, they reach thermal steam-cycle efficiencies up to 54% (i.e., net plant efficiencies of up to 43–50% on a Higher Heating Value (HHV) basis). This paper presents several possible general layouts of SCW NPPs, which are based on a regenerative-steam cycle. To increase the thermal efficiency and to match current SC-turbine parameters, the cycle also includes a single steam-reheat stage. Since these options include a nuclear steam-reheat stage, the SCWR is based on a pressure-tube design.

Topics: Water
Commentary by Dr. Valentin Fuster
2010;():279-288. doi:10.1115/ICONE18-30020.

This paper shows research and developments (R&Ds) programs for innovative technology about main components of Japan Sodium-cooled Fast Reactor (JSFR). JSFR is an advanced loop type Sodium-cooled Fast Reactor. Innovative technologies will be adopted in the JSFR for economic competitiveness, enhancing reliability, and safety. The concept of JSFR is to aim at reducing an amount of commodity, by reduction in the number of cooling loops, an adoption of high-chromium steel with low thermal expansion coefficient and high-temperature strength, and shortening a piping length by connection of outlet/inlet piping to an upper part of the reactor vessel, as well as the integration of a pump into IHX. Further, at secondary cooling system, higher reliable Steam Generator with double-walled straight tube using high-chromium steel is adopted. In the Fast Reactor Cycle Technology Development (FaCT) project, a design for JSFR has been executed along design categories such as core design, reactor system, heat transport system, safety design, etc., with corresponding R&Ds.

Commentary by Dr. Valentin Fuster
2010;():289-295. doi:10.1115/ICONE18-30023.

The High Temperature Gas-cooled Reactor (HTGR) is provided with good safety, high quality of thermal source and low cost of power generation in full life cycle. Furthermore, when the helium turbine is used for heat-work conversion, the efficiency of the HTGR is high and up to a magnitude of 50%. One of the key technologies of helium turbine is the helium compressor design. According to the conventional design rule of the air-compressor, the stage number of the helium compressor was too much excessive. Therefore, this thesis has analyzed and optimized a new cascade of helium compressor with enhanced pressure ratio in order to increase the pressure ratio and decrease the stage number. The Artificial Neural Network is used to build the approximate function which is based on database sample space. The Genetic Algorithm is used to search a new design, and the Artificial Neural Network is reused to predict the aerodynamic performance of the new design. The mean camber line and thickness distribution are optimized respectively, and the optimization results show that the total pressure loss coefficient can be reduced by 14.48% than that of the primary.

Commentary by Dr. Valentin Fuster
2010;():297-302. doi:10.1115/ICONE18-30075.

The methods of radioactive source term analysis are introduced in detail for the modular high temperature gas cooled reactor in China. Radioactive fission products and activation products produced in the reactor are described. For fission products, the emphasis is on the process from production through release to the environment for noble gas, iodine and long-lived metallic nuclides. For activation products, it mainly introduces the behaviors of H-3 and C-14. Especially the permeation process from primary circuit to secondary circuit is described for H-3. Using the preliminary design parameters of demonstration HTGR in China, basic prediction of radioactive source term is done and the results are given.

Commentary by Dr. Valentin Fuster
2010;():303-309. doi:10.1115/ICONE18-30077.

In July 2007, China entered a new era of sustainable, safe, and ecologically sound energy development by committing to build four AP1000™ units to be constructed in pairs at the coastal sites of Sanmen (Zhejiang Province) and Haiyang (Shandong Province). Both sites have the planned ability to accommodate at least six AP1000 units. The Westinghouse AP1000 is the only Generation III+ reactor to receive design certification from the U.S. Nuclear Regulatory Commission (NRC). With a design that is based on the proven performance of Westinghouse-designed pressurized water reactors (PWRs), the AP1000 is an advanced 1100 megawatt (MW) plant that uses the forces of nature and simplicity of design to enhance plant safety and operations. Excavation commenced for the first of four China AP1000 units in February 2008, and placement of the basemat concrete for Sanmen Unit 1 was completed on schedule in March 2009. This was soon followed by the placement of the first major structural module; the auxiliary building. As part of localization and the Peoples Republic of China (PRC) desire for self-reliance, a China-based module factory is constructing the major modules and manufacturing the containment vessel plates. The fabrication and welding of the containment vessel bottom head for Sanmen Unit 1 is now complete. The 2010 milestones for Sanmen Unit 1 include the setting of major modules such as the reactor vessel cavity, the steam generator, and refueling canal modules, plus containment vessel rings 1, 2, 3, and 4. All major equipment orders have been placed and the first deliveries are beginning to arrive. The technology transfer is also well underway. The Haiyang Unit 1 basemat was placed on schedule in September 2009 and Sanmen Unit 2 Nuclear Island (NI) concrete basemat placement was completed a month earlier than the milestone date of January 2010. Sanmen Unit 1 will be fully operational in November 2013 followed by Haiyang Unit 1 in May 2014. Operational dates for Sanmen Unit 2 and Haiyang Unit 2 are September 2014 and March 2015, respectively. As one of the world’s largest consumers of energy, China’s path in achieving sustainable energy has profound global economic and environmental consequences. The contract with the Westinghouse and Shaw Consortium for four AP1000 units is the largest of its type between the People’s Republic of China and the United States.

Commentary by Dr. Valentin Fuster
2010;():311-320. doi:10.1115/ICONE18-30139.

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to nine low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and the irradiations will be completed over the next five to six years to support demonstration and qualification of new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of multiple separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The second experiment (AGR-2) is currently being fabricated and assembled for insertion in the ATR in the early to mid calendar 2010. The design of test trains, the support systems and the fission product monitoring system used to monitor and control the experiment during irradiation will be discussed. In addition, the purpose and differences between the first two experiments will be compared, and updated information on the design and status of AGR-2 is provided. The preliminary irradiation results for the AGR-1 experiment are also presented.

Commentary by Dr. Valentin Fuster
2010;():321-324. doi:10.1115/ICONE18-30156.

Shimane Nuclear Power Station Unit No.3 (hereinafter NS-3) with 1,373 MW of maximum electric power being supplied to The Chugoku Electric Power Co., Inc. (Hereafter The Chugoku EPCO) is now under construction together with Hitachi-GE Nuclear Energy (HGNE). The groundbreaking (first establishment in construction business permission) was held in Dec. 2005, and is aimed at starting operation in Dec. 2011. HGNE finished installing the Reactor Pressure Vessel (RPV), the heaviest equipment in the plant in July 2009. HGNE also installed the shroud, one of the most important equipment in Aug. and a giant crawler crane was used for the installation. Regarding electrical equipment, HGNE installed a central control panel into the Control Building (C/B). The installation of main products moves ahead one after another, and the construction is going smoothly. Currently, the main construction has been proceeded to mechanical and electrical work by HGNE from building one by the building contractors. This paper reports the construction status of NS-3 and introduces applied construction technologies and construction management methods.

Commentary by Dr. Valentin Fuster
2010;():325-328. doi:10.1115/ICONE18-30163.

Electric Power Development Co., Ltd has been constructing Ohma Nuclear Power Plant aiming to start commercial operation in Nov. 2014. Ohma Nuclear Power Plant is located in Ohma-town, Aomori Prefecture and is a landmark power plant in which Mixed Oxide fuels can be loaded in the full core of the reactor. Hitachi-GE Nuclear Energy Ltd. and Kajima JV, both have extensive experience of nuclear power plant construction, are the main contractors of this project and supply the entire engineering, manufacturing of all major components, and execute the construction and commissioning for the reactor building. Ohma-town is located at the northernmost part of Aomori Prefecture bordering Tsugaru strait, where is exposed to severe cold and constant strong wind in winter. Such severe weather conditions make the construction very hard, however, Hitachi and KAJIMA tries to complete the project on schedule and on budget applying highly reliable advanced construction technologies, such as open-top and parallel construction method, all whether construction method, and large scale modularization technology. The groundbreaking (acquisition of the first construction permission) was already completed in May 2008. Its civil work steadily progressed, and the rock inspection was completed in Oct. 2009. Base mat will be completed in July 2010, and both building work and mechanical work go into full swing after installation of RCCV lower liner module.

Commentary by Dr. Valentin Fuster
2010;():329-332. doi:10.1115/ICONE18-30170.

Korea Atomic Energy Research Institute (KAERI) is developing an Advanced Power Reactor Plus (APR+) which has several improved safety features. For one of the safety features, four Emergency Core Barrel (ECB) ducts are attached at the outer surface of the reactor barrel to guide well ECC water from the Direct Vessel Injection (DVI) lines to the reactor core if necessary. The main purpose of this study is to identify the vibration characteristic of the APR+ barrel, having four ECB ducts, and to compare it with a conventional barrel. For the identification, FE analysis and vibration (modal) test for a 1/5 scale model have been carried out in air and in quiescent water. The collateral purpose of this study is to quantify the effects of fluid in a narrow gap between the barrel and the outer vessel, which are the so-called mass, damping and stiffness effects exerted by the fluid.

Commentary by Dr. Valentin Fuster
2010;():333-340. doi:10.1115/ICONE18-30172.

JAEA is now conducting “Fast Reactor Cycle Technology Development (FaCT)” project for commercialization before 2050s. A demonstration reactor for Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since FY2007 to determine referential reactor specifications for the next stage of design work of licensing and construction study. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. In this paper, the current status of the conceptual design study for the demonstration reactor plant is summarized.

Commentary by Dr. Valentin Fuster
2010;():341-348. doi:10.1115/ICONE18-30187.

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR has a large power generation capacity 1600MWe and the safety features meets European standards and regulation. This paper describes the conceptual design of EU-ABWR.

Commentary by Dr. Valentin Fuster
2010;():349-354. doi:10.1115/ICONE18-30312.

Very High temperature gas-cooled reactor (VHTR), especially the pebble-bed core type reactor, is inevitable to take place the wear of graphite components and generate the graphite dust in the core. The graphite dust was taken away by helium coolant and deposited on the surface of the primary circuit, and the fission products may be absorbed on the dust. Since it is possible that the fission products are released with dust under the accident conditions such as depressurization events, they have a potential hazard of radiation exposure to the environment. In this paper, VHTR as the research object, a testing platform is to be built with the purpose of investigating the behavior of graphite dust emission during the accident conditions. Circuit loop design is used to simulate the primary system and nitrogen is used as the working substance. Experiments of graphite dust deposition and resuspension study, as well as the study of graphite dust emission behavior during the accident conditions in pebble-bed type design VHTR will be conducted on the testing platform. The experimental data will be used for the development of VHTR source term analysis modeling.

Commentary by Dr. Valentin Fuster
2010;():355-363. doi:10.1115/ICONE18-30322.

The Japan sodium-cooled fast reactor (JSFR) is an advanced loop-type reactor concept which is investigated in the Fast Reactor Cycle Technology Development project in Japan. This paper describes some important progress on the JSFR design study and R&D discussed in a preliminary assessment for the first milestone of this project. As for the reactor system design, structural integrity against both thermal stress and seismic force was investigated. Then, the specification of the reactor vessel was established. Also, investigation of design options to extend a design margin against seismic force has been suggested. Regarding thermal hydraulics issues in the reactor vessel, design measures against cover gas entrainment and vortex cavitations have been introduced to the reactor system design based on R&D results. Further investigation is in progress for design optimization or improvement of preventive effect. The hot-leg piping design of primary cooling circuit was also established taking “Type-IV” damage into account.

Commentary by Dr. Valentin Fuster
2010;():365-368. doi:10.1115/ICONE18-30338.

The ACR-1000™ design is an evolutionary advancement of the proven CANDU® reactor design that delivers enhanced economic performance, safety, operability and maintainability. The fuel for the ACR-1000 design is based on the well established CANDU fuel bundle design that has over 40 years of demonstrated high performance. Building on its extensive experience in fuel design and analysis, and fuel testing, AECL has designed a CANFLEX-ACR™ fuel bundle that incorporates the latest improvements in CANDU fuel bundle design. The ACR-1000 fuel bundle also includes features that enable the ACR-1000 to achieve higher fuel burn-up and improved reactor core physics characteristics. To verify that the CANFLEX-ACR fuel bundle design will meet and exceed all design requirements, an extensive program of design analysis and testing is being carried out. This program rigorously evaluates the ability of the fuel design to meet all design and performance criteria and particularly those related to fuel failure limits. The design analyses address all of the phenomena that affect the fuel during its residence in the reactor core. Analysis is performed using a suite of computer codes that are used to evaluate the temperatures, deformations, stresses and strains experienced by the fuel bundle during its residence in the reactor core. These analyses take into account the impact of fuel power history and core residence time. Complementing the analyses, testing is performed to demonstrate the compatibility of the fuel with the reactor heat transport system and fuel handling systems, and to demonstrate the ability of the fuel to withstand the mechanical forces that it will experience during its residence in the core. The testing program includes direct measurement of prototype fuel element and fuel bundle properties and performance limits. A number of different test facilities are used including a cold test loop and a hot test loop with a full-scale ACR-1000 fuel channel that operates at reactor coolant temperatures, pressures and flows. This paper summarizes the out-reactor test program and related analysis that provide the basis for verifying that the ACR-1000 fuel design meets its requirements.

Topics: Fuels , Design
Commentary by Dr. Valentin Fuster
2010;():369-373. doi:10.1115/ICONE18-30343.

Atomic Energy of Canada Limited (AECL) has two CANDU® reactor products matched to markets: the Enhanced CANDU 6™ (EC6™)[1,] a modern 700 MWe class HWR design, and the Advanced CANDU Reactor™ (ACR-1000™), a 1200 MWe class Gen III+ design. Both reactor types are designed to meet both market-, and customer-driven needs. Some of the new features incorporated into the EC6 reactor include increased power output, optimized maintenance outages, more automated testing and an Advanced Control Room. Lessons learned through feedback obtained from the operating plants have been incorporated into the design, and equipment obsolescence has been addressed. This paper presents basic EC6 design improvements; AECL works with its customers to assess their individual design requirements. Excellent neutron economy, on-power refueling, a simple fuel bundle, and the fundamental CANDU fuel channel design provide the EC6 reactor with unsurpassed flexibility in accommodating a wide range of advanced fuels and fuel cycles in addition to the standard natural uranium. These advanced fuels provide the promise of extending resources, reducing waste and enhancing proliferation resistance.

Topics: Fuels , Cycles
Commentary by Dr. Valentin Fuster

Codes, Standards, Licensing and Regulatory Issues

2010;():375-383. doi:10.1115/ICONE18-29049.

In this paper, we address fatigue verification of Class 1 nuclear power piping according to ASME Boiler & Pressure Vessel Code Section III (ASME III), NB-3600, and several relevant issues that are often discussed in connection to the power uprate of several Swedish BWR reactors in recent years. We review first the basic requirements and their verifications using finite element analysis in detail. Thereafter, we clarify a so-called simplified elastic-plastic discontinuity analysis for further verification if the basic requirements found unsatisfactory, and examine necessary computational procedures for evaluating alternating stress intensities and cumulative damage factors. Our emphasis is placed on alternative verification procedures, which do not violate the general design principles upon which ASME III NB-3600 is built, when fatigue damage usages predicted by the simplified elastic-plastic discontinuity analysis are unaccepted. An alternative which employs a non-linear finite element computation and a refined numerical approach for re-evaluating the cumulative damage factors is suggested. Concluding remarks are given.

Commentary by Dr. Valentin Fuster
2010;():385-394. doi:10.1115/ICONE18-29276.

The present paper describes a new method for determining the target value of structural reliability in the framework of the System Based Code (SBC) by considering the safety point of view. In the new method, the reliability target is derived from the proposal to a quantitative safety goal that was published by the nuclear safety commission (NSC) of Japan and the quantitative safety design requirements on the core damage frequency (CDF) and the containment failure frequency (CFF) that were determined in the Fast Reactor Cycle Technology Development (FaCT) project by Japan Atomic Energy Agency (JAEA), by utilizing analysis models of a probabilistic safety assessment (PSA). The present method was applied to determination of the reliability target of the structures and components which constitute the reactor cooling system in the Japanese sodium-cooled fast reactor (JSFR). The risk from the reactor is expressed with sum of combination of various elements in the PSA analysis model. Those elements include not only static failure of the structures and components. However, the present study focuses on the sequences including the static failure, and the probability of dynamic failures and human errors in those sequences is conservatively assumed as a unity. It was confirmed that the present method combined with the PSA analysis model for internal initiating events is applicable to determination of the reliability target associated with a random failure of the structures and components, and that the method related to seismic initiating events can derive the target value of the occurrence frequency at which any of the important structures and components fails due to an earthquake.

Topics: Reliability
Commentary by Dr. Valentin Fuster
2010;():395-401. doi:10.1115/ICONE18-29330.

Nuclear power generation has become an increasingly attractive alternative in the United States (U.S.) power market due to several factors: growing demand for electric power, increasing global competition for fossil fuels, concern over greenhouse gas emissions and their potential impact on climate change, and the desire for energy independence. Assuring the protection of people and the environment are of paramount concern to nuclear power generators and regulators as we move towards a possible nuclear renaissance. Thus, sound engineering design is of utmost important and potential environmental and safety concerns must be carefully evaluated and disposition during permitting of the new nuclear power plants. Areas to be considered in order to alleviate these concerns include the following: • Site meteorology and dispersion conditions of the area; • Evaluation of radiological consequence during normal plant operation and emergency conditions; • Water availability for plant cooling system; • Evaluation of potential land use, water use, ecological and socioeconomic impacts of the proposed action. This paper focuses on site suitability evaluation for greenfield sites through site characterization, examination of challenges/constraints in deployment of available technology/plant systems, and mapping of permitting compliance strategy. Case studies related to selection of plant systems based on the environmental site conditions, preferred compliance plan, and public acceptance, are included.

Commentary by Dr. Valentin Fuster
2010;():403-407. doi:10.1115/ICONE18-29331.

Nuclear power generation has become an increasingly attractive alternative in the global power market due to growing demand for electric power, increasing global competition for fossil fuels, concern over greenhouse gas emissions and their potential impact on climate change, and the desire for energy independence. Nuclear energy plays an integral role in providing carbon free energy for sustainable development of global electric power generation. Assuring the protection of people and the environmental is a prime consideration in the design, construction, and operation of nuclear power plants. Potential environmental and safety concerns must be carefully evaluated and addressed. In order to assure that the nuclear power plant designs are sufficiently robust, the U.S. Nuclear Regulatory Commission (USNRC) requires that applicants for early site permits (ESP) and construction/operating licenses (COL) identify the most severe natural phenomena historically reported for the site and surrounding area to ensure sufficient design margin exists, considering the limited accuracy, quantity, and time in which the associated data has been collected. Because these permits are valid for a period up to 40 years, the potential impacts of climate change on the severity of natural phenomena, as it relates to the design basis and nuclear safety and environmental impacts are of increasing interest. Although no conclusive evidence or consensus of opinion is available on the long-term climatic changes resulting from human or natural causes, the USNRC has requested that climate change forecasts be considered for their potential affecting the most severe natural phenomena. The specific subject areas of concern include: • Extreme temperature and extreme precipitation (liquid & frozen) statistics – review 100 years of data around the site versus a review of the previous 30 years of data. • Extreme wind/basic wind speed – review previous 100 years of tropical cyclone data (including hurricanes) in the site vicinity versus previous 30 years of data. • Tornado – review of frequency and intensity trends and forecasts. • Drought – review of water availability / water supply during drought conditions and drought of record. • Stagnation Potential – review of conditions that would result in restrictive dispersion of greenhouse gas emissions. This paper examines the challenges and constraints in identifying and developing appropriate design- and operating-basis site/regional meteorological conditions while accounting for potential climate change during preparation of an ESP and/or COL. Because there is no regulatory guidance or quantitative acceptance criteria currently available, the methodology used to address climate change in a recent issued ESP will be discussed as an example.

Commentary by Dr. Valentin Fuster
2010;():409-412. doi:10.1115/ICONE18-29406.

Nearly about 100 nuclear power national standards and 340 industry standards have been made in the field of nuclear power in China, which can partly meet the need of development of the second generation pressurized water reactor nuclear power plant. The situation of nuclear power fast development in China causes more standards to be made in the field of nuclear power, furthermore, the standard system and standardization have some problems now, to make more standards scientifically and solve existing problems, a number of measures should be taken in future.

Topics: China , Nuclear power
Commentary by Dr. Valentin Fuster
2010;():413-418. doi:10.1115/ICONE18-29657.

The System Based Code (SBC) concept has been proposed to achieve compatibility in matters of reliability, safety, and cost of Fast Breeder Reactors (FBR). This code extends the present structural design standard to include the areas of load setting, fabrication, inspection, maintenance, and so on. Therefore, a quantitative index which can connect different areas is required. In addition, the determination of its target value is also one of the key points to substantiate the SBC concept. Failure probability is one of the candidate indexes. We have proposed a new method to determine the reliability targets for the structures and components in FBR plants from the safety point of view by utilizing analysis models of a probabilistic safety assessment. In this study, the effectiveness of the failure probability as an index and the compatibility of the reliability targets derived by the new method were investigated through a trial setting of In-Service Inspection (ISI) request on the reactor vessel near the sodium surface based on the SBC concept. The failure probability due to creep-fatigue interaction was calculated by the Monte-Carlo simulation. In response, the reliability targets for fracture related to the risk from internal initiating events were derived. Cumulating the failure probability and the reliability targets up to the end of in-service period enables us to compare them directly, and we obtained a result that the reactor vessel has enough reliability even without ISI. Through this trial, we showed that the failure probability is a promising index, and the reliability targets derived by the new method are compatible with the SBC concept.

Topics: Reliability
Commentary by Dr. Valentin Fuster
2010;():419-423. doi:10.1115/ICONE18-29752.

The American Nuclear Society is developing a standard titled “Low Power and Shutdown PRA Methodology, ANSI/ANS-58.22.” It has been under development for more than 10 years now. During this time, drafts of the standard have been balloted by the ANS American Nuclear Society (ANS) Risk-Informed Safety Committee (RISC) on three different occasions; i.e., in 2005, 2006, and 2008. The most recent of these ballots occurred in November 2008; i.e., Reference 1. Each ballot failed to achieve a consensus for approval and numerous comments were received for improvement. Since the completion of the ballot on Version 8C in 2008, the Low Power and Shutdown (LPSD) writing group has been working on a volunteer basis to respond to the approximately 600 comments received and to revise the Standard for future ballot. A revised standard was completed in early November 2009. This latest draft was submitted in November 2009 to the ANS RISC pending a formal ballot for informal review. The intention of this informal review is to ascertain if members of the balloting RISC committee could support the revised standard and if not, to identify those issues for further work by the writing group. For approval of ANS standards, a consensus for approval must be received among all stakeholder groups, rather than a simple majority to approve.

Topics: Safety
Commentary by Dr. Valentin Fuster
2010;():425-431. doi:10.1115/ICONE18-29773.

The objectives of this paper is to discuss technical harmonization of Nuclear Codes and Standards, based on French long experience in Codes and Standards used for design-fabrication and operation of nuclear components (mainly pressure retaining components). After a long period of use of ASME Section III code, during the Westinghouse licensing process, AFCEN (AREVA, EDF and the major manufacturers) decided to develop their own AFCEN French Codes. The 1st version has been issued in 1980 and the last one in 2007, completed by annual addendum. During more than 20 years the 2 Codes, RCCM and ASME Section III, have leave separately, with different constraints like industrial history, localisation of fabrication, more new plants in France than in USA, different R&D programs to support Code improvement[[ellipsis]] Recently a detailed review of differences for class 1 vessel has showed under a “general global quality equivalence”, a lot of differences in the Code development process, in the Code organization, in the scopes, in the State of the Art fulfillment, in ageing consideration at the design stage, in relation with national or international regulations, in term of standards used or complementary specification needs[[ellipsis]] The harmonization of Codes and Standards is possible under an important effort to move toward new ideas, more international rules and with a strong support of national safety authorities.

Commentary by Dr. Valentin Fuster
2010;():433-437. doi:10.1115/ICONE18-29806.

The EUR organization was set up in the early 90’s to produce a common specification for the next LWR nuclear power plants to be built in Europe. 18 years after its foundation, the organization is well recognized and still quite active. During the last 3 years, the EUR products have mainly been evaluations of the Gen 3 LWR designs. The evaluations of the AP1000 and of the AES92 designs have been concluded in 2007 and a revised version of the evaluation of the EPR completed in 2009. Other LWR projects of potential interest for the EUR utilities, such as MHI’s APWR, are being reviewed before starting a full-scope assessment. Last, a revision C of the EUR volume 4 has been published in 2007. Coordinated actions with the other industry groups and the other stakeholders have been a centerpiece of the recent EUR strategy. In particular, the EUR and ENISS organizations have joined their efforts in nuclear safety vs. IAEA and WENRA. Also EUR and WNA/CORDEL are now working together on harmonization of the design requirements at global level. Meanwhile, the EUR organization has kept enlarging: CEZ and MVM now are active associated members and Gen-Energija from Slovenia has been invited to participate. Education and training has been dealt with actively in 2009 and 2010. The EUR organization strongly supported WNU’s “Forum on harmonization” in 2009. A more technical course about the EUR requirements is being organized in 2010 under the aegis of ENEN. Finally, a lot of preparatory material for a revision D of the EUR volumes 1 and 2 has been gathered during the last 10 years. Several options are under consideration about how to proceed towards this revision.

Commentary by Dr. Valentin Fuster
2010;():439-442. doi:10.1115/ICONE18-29852.

Fluor, a large Engineering, Procurement and Construction contractor, recently renewed its ASME certifications for the construction of nuclear power plants. In preparation for a resurgence of commercial nuclear power plant construction, Fluor Nuclear Power (FNP) Construction Welding Engineers have prepared an electronic field welding program, to be used in conjunction with an automated system for the generation, control and documentation of work packages. The prior generation of nuclear power plants constructed in the US utilized a manual process for controlling field welding activities. The “manual” way of doing business required a relatively large, on-site staff (both technical and administrative) to create, issue, track, and document this work. In addition, the manual process was prone to human error. In an effort to improve this key construction activity and reduce construction costs, the FNP Construction Welding group has prepared an electronic welding program that automatically performs the majority of the work package preparation/documentation tasks previously performed manually. The electronic welding system has been designed to access engineering and construction code information related to welding, process the data through a series of logic-based spreadsheets and automatically populate the work package with welding requirements — preheat, post-weld heat treatment (PWHT), Welding Procedure Specification (WPS), etc. The spreadsheets analyze the engineering data (i.e., base material type, thickness, applicable code, joint design, etc.), in conjunction with construction code rules and Fluor welding practices, to determine appropriate welding requirements. System generated requirements are then automatically entered into the work package. This paper describes the design of the electronic welding program, it’s scope, development and qualification. In addition, preparation and qualification of the spreadsheet logic, that effectively translates specified code welding criteria into work package requirements, will be reviewed. It is believed that this type of system will be needed to successfully construct the next generation of US nuclear power plants.

Topics: Welding , Construction
Commentary by Dr. Valentin Fuster
2010;():443-448. doi:10.1115/ICONE18-29939.

The present Chinese nuclear program is developing vastly and rapidly. The main type of nuclear units under construction currently is the improved generation II PWR (G2+). Taking into account that G2+ is designed based on French M310 type, and one of the technical bases and guidance that NNSA uses to evaluate G2+ is the French RCC codes, AFCEN has authorized CNPRI to translate RCC codes into Chinese version and publish them, in order to facilitate Chinese manufacturers to better apply RCC codes. CNPRI organizes several sub-groups to perform translation, revise, publication and coordination working groups, meanwhile AFCEN provides technical support. Translation is realized based on RCC codes English version, with some reference of French version. Translators work with professional technical background and knowledge, in case of non-clear understanding, including language and technical issues, they raise question list to AFCEN experts for clarification, afterward the answers and comments are merged accordingly into Chinese version translation, to assure the accuration, better understanding and the guidance to Chinese G2+ program.

Commentary by Dr. Valentin Fuster
2010;():449-455. doi:10.1115/ICONE18-30115.

Today, commercial nuclear power plants are installing High-Density Polyethylene (HDPE) piping in non-safety-related and safety-related systems. HDPE has been chosen in limited quantity to replace carbon steel piping as it does not support rust, rot, or biological growth. However, due to its relatively short nuclear service history, developing a way to accurately evaluate joint integrity in HDPE is crucial to utilities and the U.S. Nuclear Regulatory Commission (USNRC). This paper will investigate using ultrasonic Phased Array to quantify detection of flaws and detrimental conditions in butt-fusion joints throughout the full spectrum of applicable HDPE pipe diameters and wall-thicknesses. Currently the most concerning joint condition is that of “Cold Fusion”. A cold-fused joint is created when molecules along the fusion line do not fully entangle or co-crystallize. Once the fusion process is complete, there is the appearance of a good, quality joint. However, the fabricated joint does not have the required strength as the co-crystallization along the pipe faces has not occurred. Therefore, performing a visual examination of the bead, as required by the current revision of ASME Code Case N-755, does not provide adequate assurance of joint integrity. As a potential solution, volumetric examination is being considered by the USNRC to safeguard against this and other types of detrimental conditions. Factors addressed will include pipe diameter, wall-thickness, fusing temperature, interfacial pressure, dwell (open/close) time, and destructive correlation with ultrasonic data.

Topics: Density
Commentary by Dr. Valentin Fuster
2010;():457-462. doi:10.1115/ICONE18-30243.

With the resurgence of nuclear power and increased interest in advanced nuclear reactors as an option to supply abundant energy without the associated greenhouse gas emissions of the more conventional fossil fuel energy sources, there is a need to establish internationally recognized standards for the verification and validation (V&V) of software used to calculate the thermal-hydraulic behavior of advanced reactor designs for both normal operation and hypothetical accident conditions. To address this need, ASME (American Society of Mechanical Engineers) Standards and Certification has established the V&V 30 Committee, under the jurisdiction of the V&V Standards Committee, to develop a consensus standard for verification and validation of software used for design and analysis of advanced reactor systems. The initial focus of this committee will be on the V&V of system analysis and computational fluid dynamics (CFD) software for nuclear applications. To limit the scope of the effort, the committee will further limit its focus to software to be used in the licensing of High-Temperature Gas-Cooled Reactors. In this framework, the Standard should conform to Nuclear Regulatory Commission (NRC) and other regulatory practices, procedures and methods for licensing of nuclear power plants as embodied in the United States (U.S.) Code of Federal Regulations and other pertinent documents such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods” and NUREG-0800, “NRC Standard Review Plan”. In addition, the Standard should be consistent with applicable sections of ASME NQA-1-2008 “Quality Assurance Requirements for Nuclear Facility Applications (QA)”. This paper describes the general requirements for the proposed V&V 30 Standard, which includes; (a) applicable NRC and other regulatory requirements for defining the operational and accident domain of a nuclear system that must be considered if the system is to be licensed, (b) the corresponding calculation domain of the software that should encompass the nuclear operational and accident domain to be used to study the system behavior for licensing purposes, (c) the definition of the scaled experimental data set required to provide the basis for validating the software, (d) the ensemble of experimental data sets required to populate the validation matrix for the software in question, and (e) the practices and procedures to be used when applying a validation standard. Although this initial effort will focus on software for licensing of High-Temperature Gas-Cooled Reactors, it is anticipated that the practices and procedures developed for this Standard can eventually be extended to other nuclear and non-nuclear applications.

Commentary by Dr. Valentin Fuster
2010;():463-467. doi:10.1115/ICONE18-30256.

VERLIFE — “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for WWER type NPPs, as these codes were developed only for design and manufacturing and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. To assure that VERLIFE Procedure will remain a living document, new 3-years IAEA project (in close co-operation with the another project 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) has started in 2009. Within this project, upgrading/updating of the VERLIFE procedure is prepared together with the extension by (at least) following procedures: - Leak-before-break concept for WWER NPPs; - Reduction of Probability of Break procedure for evaluation of integrity of high-energy piping in NPPs of WWER-440 and WWER-1000 types; - Lifetime of reactor pressure vessel internals; - Risk informed In-service inspection implementation process and organization; - Methodology for Qualification of In-Service Inspection Systems for WWER Nuclear Power Plants; - Component and piping supports; - Monitoring and evaluation of erosion-corrosion damage in piping materials. Final document, after its approval by expert groups of the IAEA and NULIFE, will be issued as “IAEA/NULIFE Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs”. The paper will describe these main principles and also future plans.

Topics: Pipes
Commentary by Dr. Valentin Fuster
2010;():469-477. doi:10.1115/ICONE18-30260.

This paper provides a follow up to prior works [1,2,3,4] describing the theory and method associated with an innovative microwave based NDE method for inspection of HDPE thermal and electro fusion pipe joints. The method employs a first of a kind apparatus that is based on the creation of an image using electromagnetic energy in the microwave frequency range. This paper presents the results of a series of field and laboratory trials of the HDPE thermal and electro-fusion inspections and the corresponding mechanical test of the joint. A close correlation between the inspection results and the mechanical test results are shown. The data presented provides definitive evidence of the ability of the inspection method to detect numerous typical defect types in HDPE fusion joints.

Commentary by Dr. Valentin Fuster
2010;():479-483. doi:10.1115/ICONE18-30295.

Analyses were performed by Kalsi Engineering, Inc. (KEI) on the main steam isolation valves (MSIV) and main steam check valves (MSCV) at 4 pressurized water reactors to evaluate their capability to perform design basis functions at Extended Power Uprate (EPU) conditions. Disc closing velocities and actuator pressures (for MSIV) were calculated by using RELAP™ . These inputs included valve specific flow and torque coefficients obtained by detailed Computational Fluid Dynamics (CFD) analyses to accurately predict hydrodynamic torque imposed on the disc throughout the stroke. Stresses, strains and displacements in all of the critical MSCV and MSIV components were determined by performing appropriate 2-D and 3-D elasto-plastic, transient dynamic finite element analysis (FEA). Fatigue life of the components under spurious closure was determined in accordance with applicable ASME Section III Code. The evaluation also included the propensity for excessive wear of the MSCV due to disc flutter, vibration and erosion.

Commentary by Dr. Valentin Fuster
2010;():485-487. doi:10.1115/ICONE18-30346.

Use of the engineering, procurement and construction (EPC) contracting arrangement is increasing in China recently, especially for those important construction projects. The nuclear power plant construction projects are as well. Because of the associated hazards, nuclear power plant construction projects require more strict safety and quality requirements as compared with more conventional construction projects. As essential means and safety shields, the equipment quality supervision and surveillance requirements and their implementation mode should be explored to meet the requirements of EPC projects. It is highly demanded to establish new rules to meet the quick development of nuclear power plant construction projects. The article discusses the current equipment quality supervision and surveillance mode first, and then discusses the needs to build rules especially the significance of introducing independent organizations to supervise equipment quality for those nuclear power plant construction projects for using EPC contracting arrangement despite the supervision and surveillance conducted by manufactures, EPC contractors, nuclear power plant owners and the National Nuclear Safety Administration.

Commentary by Dr. Valentin Fuster

Fusion Engineering

2010;():489-494. doi:10.1115/ICONE18-29030.

The ITER is an international collaborative project aimed at demonstrating the scientific and technological feasibility of fusion energy for peaceful purposes. China as one of the seven parties takes part in the ITER, and wishes to grasp the remote handling technology, which is one of the four key technologies related to the future fusion reactors for electric power generation. The transfer cask system (TCS) is one subsystem of ITER remote handling system, which provides the means for the remote transfer of (clean/activated/contaminated) in-vessel components and Remote Handling Equipment between Hot Cell Facility and Vacuum Vessel through dedicated galleries and lift in the ITER buildings. The TCS can work in the nuclear radiation environment and can be fully driven by self powered electricity with high energy density batteries. Its driving force is provided by nearly twenty servo motors. The remote handling technology can lay the foundation for developing demonstration nuclear fusion power plant in China on self-reliance. Due to the gamma irradiation and the hazard material in these ITER parts, all required maintenance of the port plug and the inner components are being carried out by the TCS, which offers confinement boundaries to these components. The ITER Tokamak building includes three floors, including upper port level, equatorial port level and lower port level, linked by a lift. Due to limited Tokamak building space which is frozen and can not be changed presently, the TCS penetrates its cable tray for about 300 mm. According to the configuration each port level and the mass of the corresponding plug, the dimensions of the TCS envelope in three levels are different. The basic components and the basic parameters of the TCS are presented. Furthermore, according to each port level configuration and the safety requirement of the TCS, the radius of the curvature with the TCS trajectory is optimized, and a trajectory of each port level is determined by the positioned guidance beacons. At last, the results of the computer aided design (CAD) shows that the present conflict between TCS and Tokamak building can be designed compatible with the proposed variable structure cable tray in the ITER Tokamak building and the TCS based on a fleet of server motor driven system.

Commentary by Dr. Valentin Fuster
2010;():495-499. doi:10.1115/ICONE18-29086.

The ITER (International Thermonuclear Experimental Reactor) is an international collaborative project, and its object is aiming at demonstrating the scientific and technological feasibility of fusion energy for peaceful purposes. In the ITER, the four important engineering challenges are the first wall of the blank model, the remote handling (RH), the heating of the plasma, and the superconducting technology. The RH control system is very complexity, and the total number of the control nodes is 15378, which consists of the control nodes in the Tokamak equipment, the hot cell equipment and the control room equipment. Only the number of the control nodes of the transfer cask system (TCS) is up to 2580. The TCS is one sub system of ITER RH system, which provides the means for the remote transfer of (clean/activated/contaminated) in-vessel components and remote handling equipment between hot cell facility and vacuum vessel through dedicated galleries and lift in the ITER buildings. Due to the experimental facility with a very long timeline, the control software on the ITER is better suited using an open source solution as compared to a commercial solution. According to market share and proven record, command control and data acquisition and communication (CODAC) group chooses the EPICS (Experimental Physics and Industrial Control System) as the preferred solution. China takes part in the ITER, and wishes to grasp the RH technology, which is one of the four key technologies related to the future fusion reactors for electric power generation. According to the CODAC system configuration with EPICS component, the concept control architecture of the TCS is presented in this paper. The control system consists of the file network, the interlock network, the safety network, real-time network, RH control network and visual supervising network etc. Then, the paper analyses each part corresponding requirement function. As a result, a development service platform concept is set up. Finally, the experiment system of the leg of the TCS has been constructed.

Commentary by Dr. Valentin Fuster
2010;():501-507. doi:10.1115/ICONE18-29117.

The moving direction of double seal door (DSD) of ITER remote handling transfer cask and the force of hydraulic pole will change significantly at the guide rail inflexion position (GRIP) which is a mutant site, so it is very possible to make the structure damage or the system failure at the GRIP. In this paper, the kinematics simulation and analysis of DSD were done based on special constitution restriction and working process by software ADAMS. The stress distribution of guide rail and hydraulic pole were obtained by the above simulation, at the same time the optimal GRIP was confirmed according to the force analysis result. The above-mentioned analysis process and results not only provide technical data for the optimization design and the prototype manufacture of DSD, but also provide the examples and references of kinematics analysis for other important components of ITER.

Topics: Kinematics , Doors
Commentary by Dr. Valentin Fuster
2010;():509-515. doi:10.1115/ICONE18-29146.

Fusion energy with much more advantages could become a serious option for the future energy. The EAST (Experimental Advanced Superconducting Tokomak) device is aiming to study the physical and technical issues involved in stead-state tokamak nuclear fusion. Cryogenic system is one of the most important sub-systems in EAST device, and turbine-expanders are the key components of cryogenic system. In this paper, turbine-expanders with specific configuration in EAST cryogenic system were introduced, and the problem of instability of turbine-expanders in actual refrigerating process was pointed out. Three different refrigerating methods (Using T3, No-using T3 and using a two-phase turbine) were analyzed, and the results showed that the instability of turbine-expander, especially the coldest turbine (T3), will influence the normal operation, even the whole physical experiment of EAST device. The liquefaction capacity will be increased largely by using a two-phase turbine-expander instead of the JT-throttling.

Topics: Bearings , Turbines , Helium
Commentary by Dr. Valentin Fuster
2010;():517-522. doi:10.1115/ICONE18-29262.

Fire-extinguishing behavior of four fire extinguishants, dry sand, pearlite, Natrex-L and Natrex-M on burning lithium was examined. Temperature and flame increase in chemical reaction between lithium and silicon, which is the major element in the fire extinguishants, were observed for dry sand and pearlite. For Natrex-L, temperature increase was not observed visually, although flame was slightly increased when it was applied to the burning lithium. The effect of lithium pool depth on the fire-extinguishing performance of Natrex-L was investigated on the definite area of the lithium combustion surface because the density of Natrex-L was larger than that of liquid lithium. It was found that the amount (thickness) of fire extinguishant necessary for fire-extinguishing increased as the depth increased. In this experimental condition (combustion area: 270cm2 , lithium depth: 1–2cm), the minimum thickness of the fire extinguishant was 1.5 times the depth of the lithium pool.

Topics: Fire , Lithium
Commentary by Dr. Valentin Fuster
2010;():523-531. doi:10.1115/ICONE18-29268.

The International Fusion Materials Irradiation Facility (IFMIF) is a D+ -Li neutron source aimed at producing an intense high energy neutron flux (2 MW/m2 ) for testing candidate fusion reactor materials. Under Broader Approach activities, Engineering Validation and Engineering Design Activities (EVEDA) of IFMIF started on July 2007. Regarding to the lithium (Li) target facility, design and construction of EVEDA Li Test Loop is a major activity and is in progress. This paper presents the current status of the design and construction of EVEDA Li Test Loop. The EVEDA Li Test Loop consists of a main loop system and a purification loop system. The detail design was started at the early 2009. Fabrication of the loop was started at middle of 2009, and completion is planned at the end of Feb. 2011. Currently, the system diagram of the EVEDA Li Test Loop is finished to be defined. The diagram and function of major components in the main loop system and the purification loop system are described in this paper.

Commentary by Dr. Valentin Fuster
2010;():533-537. doi:10.1115/ICONE18-29367.

For EAST operation schedule, auxiliary subsystems like the cryogenic system have to cope with different heat loads which depend on the different EAST operating states. The cryogenic system consists of a cryoplant and a cryodistribution system. All of these cryogenic subsystems have to operate in parallel to remove the heat loads from the magnet, 80K shields, built-in cryopumps and other small users. After a brief recall of the main particularities of a cryogenic system operating in a Tokamak environment, the first part of this study is dedicated to the assessment of the main EAST operation states. A new design of refrigeration loop for the HTS current leads, the updated layout of the cryodistribution system and revised strategy for operations of the built-in cryopumps have been taken into consideration. The relevant normal operating scenarios of the cryoplant are checked for the typical EAST operating states like plasma operation state, short term stand by, short term maintenance, or test and conditioning state. The second part of the paper is dedicated to the abnormal operating modes coming from the magnets and from those generated by the cryoplant itself. Thanks to this analysis, the optimization of the present operational modes is proposed to make match the technical specifications of the cryogenic system with the EAST operation requirements.

Commentary by Dr. Valentin Fuster
2010;():539-546. doi:10.1115/ICONE18-29427.

The development of the conceptual design of the IFMIF Target and Test Cell (TTC) is briefly summarized by outlining the previous reference TTC design and the current Modular TTC (MTC) concept. Based on the MTC concept, the latest progresses of the preliminary engineering design of the key TTC components, including the TTC vessel, the Top Shielding Plugs (TSPs), the Removable Intermediate Ring (RIR), and the Test Module Interface Head (TMIH), are described. A specimen flow, based on handling requirements of the High Flux Test Module (HFTM), between the TTC, Access Cell (AC), Test Module Handling Cell (TMHC) and the Post Irradiation Examination (PIE) facilities is proposed as well as the function of the AC and TMHC is preliminary defined. The TMHC is proposed to be divided into a Component Handling Cell (CHC) and a Rig Handling Cell (RHC) regarding the dimension differences of the components to be handled inside of the cells. The recycling of the irradiated specimens for another campaign of irradiation is also considered in this specimen flow.

Commentary by Dr. Valentin Fuster
2010;():547-552. doi:10.1115/ICONE18-29534.

Decrease of pores in tritium permeation barriers is one of the most important problems to be addressed for the proper functioning of the fusion reactor. In this paper, a self-healing composite coating composed of TiC+mixture (TiC/Al2 O3 ) +Al2 O3 was developed to solve this problem. The coating was deposited on martensitic steels by plasma spraying with a thickness of 100μm. After heat-treatment, the morphology and phase of the coating were investigated by scanning electron microscopy (SEM) and X-ray diffraction (XRD). The corrosion resistance of the coating before and after heat treatment was examined by electrochemistry techniques. The results showed that the TiC+mixture (TiC/Al2 O3 )+Al2 O3 coating exhibited good adhesion to the substrate and a perfect self-healing ability with the porosity decreased by 90% after heat-treatment. The corrosion resistance of the coating increased evidently after the heat treatment. The oxidation/expansion of TiC in the coating played an important role in the sealing of pores.

Commentary by Dr. Valentin Fuster
2010;():553-557. doi:10.1115/ICONE18-29586.

Based on the structural design of the Chinese ITER Dual Functional Lithium-Lead Test Blanket Module (DFLL-TBM), Three Unprotected Loss of Flow Accidents (ULOFAs) were investigated preliminarily, assuming that the whole nuclear heat in TBM was carried away by the flowing lithium-lead (LiPb). The results show that the temperature of the first wall (FW) increases rapidly and the maximum temperature appears at the lower part of FW. In the analysis of ULOFAs, the maximum temperature might exceed the melting point of structure material steel. This event must be avoided by the fusion power shutdown system that terminates plasma burn.

Topics: Lithium
Commentary by Dr. Valentin Fuster
2010;():559-564. doi:10.1115/ICONE18-29625.

International Thermonuclear Experimental Reactor (ITER) TF feeder systems convey the cryogenic supply and electrical power to the TF coils. The Cryostat Feed-through (CFT) includes the straight feeder part from the cryostat wall to the S-Bend Box (SBB). It is the bottleneck of the feeders. The huge Lorentz-force is a challenge for the CFT design. So the reasonable distribution and structural design of the internal and external supports are important. The CFT include the cold (cryogenic) to warm (room temperature) transitions. It is highly integrated with the cryo-pipes, the busbars, the superconductor joints, the thermal radiation shield and the instrumentation pipes and so on. The cryogenic and electrical requirements, the vacuum and mechanical requirements, and so on are considered when the CFT is designed. This paper presents the functional requirements on the TF CFT, gives its structure. The supports are designed and arrayed according to their mechanical or thermal function separately to stand the huge mechanical loads and isolate the conducting heat load from room temperature respectively. The assembly scheme is also described. Mid-joint and cryostat joint are designed to give the facility for the assembly on location. The mechanical analysis result shows the stress in the stainless steel and G10 material both are within the materials stress safety margin. The heat load to the cryogenic pipes and busbars are also less than the requirement 15W. Transient thermal analysis of global feeder model indicates that 32 days are needed for the feeder components to cool down to the required condition.

Topics: Magnets , Design
Commentary by Dr. Valentin Fuster
2010;():565-571. doi:10.1115/ICONE18-29634.

This paper reports on the measurement of surface waves on a liquid lithium jet and results of the study of the Li target at the International Fusion Materials Irradiation Facility (IFMIF). The characteristics of the surface waves at the nozzle exit and just downstream of it were examined experimentally, since the initial growth of free surface waves exerts a definite influence on surface behavior in the downstream region. Experiments were carried out using the lithium circulation loop at Osaka University, with a focus on the free surface oscillations. These oscillations were measured using an electrocontact probe apparatus, which detects electric contacts between the probe tip and the Li surface. The apparatus was installed 15 mm downstream from the nozzle exit and was scanned along the liquid-depth direction. The contact signal recorded in the experiment was analyzed, and the wave amplitude and frequency of the surface waves were examined.

Commentary by Dr. Valentin Fuster
2010;():573-578. doi:10.1115/ICONE18-29743.

In inertial confinement fusion (ICF) project, there are many diagnostic methods. Neutron penumbral imaging is one of the important technologies to diagnose the information about neutron spatial and temporal distribution in burn region of the core of a compressed pellet in the low yield fusion. In this study, a linear space invariant neutron penumbral imaging system was designed and established with Monte Carlo method, the system’s point spread function (PSF) was obtained. By fitting the obtained PSF, several mathematic models were obtained and compared. The improved “logistic function” mathematic model was chosen to reconstruct the coded penumbral image and the original neutron source image “T” was successfully was obtained.

Topics: Optimization , Imaging
Commentary by Dr. Valentin Fuster
2010;():579-584. doi:10.1115/ICONE18-29747.

Alumina has been identified as a most promising material in tritium permeation barrier. By introducing some additive into the alumina improves its effectiveness. The purpose of this study was to synthesize Al2 O3 coated SiC particles by heterogeneous precipitation method to replace the pure alumina as the raw material of coating. A series of factors which influenced the properties of coated powders had been studied, including the concentration of Al(NO3 )3 solution, surface active reagent, the addition amount of SiC, calcinations temperature etc. The coated powders were analyzed by scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), X-ray diffraction (XRD) and Zeta-potential analyzer. The result indicates that the addition of certain amount of nano powers can reduce the phase transformation temperature of the alpha alumina. α-Al2 O3 coated SiC powders were obtained by calcining Al(OH)3 /SiC precursor at 1100 degree centigrade.

Topics: Precipitation
Commentary by Dr. Valentin Fuster
2010;():585-589. doi:10.1115/ICONE18-29795.

A multijunction antenna unit has been designed and will be used to fabricate the new 4.6GHz Lower Hybrid Current Drive (LHCD) launcher for the long pulse operation of the Experimental Advanced Superconducting Tokamak (EAST). In this paper, the construction and design method of the multijunction antenna unit are described, and then the reflection coefficients of the unit with the matched load and with the plasma load are analyzed and discussed respectively by using High Frequency Structure Simulator (HFSS) code and by using numerical calculation based on linear coupling theory. The simulation result shows that the return loss of the antenna unit keeps below −25dB in a bandwidth of 100MHz and the power is distributed equally in the secondary waveguides of the unit. Numerical calculation result also shows that the unit has good coupling property in a wide range of tokamak plasma parameters.

Commentary by Dr. Valentin Fuster
2010;():591-596. doi:10.1115/ICONE18-29901.

The Poloidal Field (PF) coils, made of superconducting Cable in Conduit Conductor (CICC), are important for keeping plasma inside the vacuum vessel of ITER Tokamak device. To ensure the safety of the PF coil, Non-Destructive Testing (NDT) has to be performed for each conduit tube of CICC, as it is the boundary of liquid Helium and supporting structure of the electromagnetic load. However, the complicated cross section shape of the PF conduit tube causes difficulty in its NDT. Aiming at solving the NDT problem of the PF conduit tube, two approaches are proposed and their validity is demonstrated in this paper. One is a hybrid strategy of the ultrasonic phased array and the eddy current array technology. The other is of the immersed ultrasonic testing method. The experimental results reveal that both approaches are promising for the NDT of ITER conduit tubes though the later one is more efficient.

Topics: Cables
Commentary by Dr. Valentin Fuster
2010;():597-604. doi:10.1115/ICONE18-29912.

The ITER project is basically an engineering and construction project in order to build the ITER machine which is a scientific experimental fusion device. The seven members of the project have all created legal entities called Domestic Agencies to provide in-kind contributions to the ITER Organization (IO) for the supply of components which are manufactured by their suppliers. According to ITER agreement and due to nuclear safety involved in the fusion process, the project requires a license from the French Nuclear Safety Authority. One of nuclear safety regulations is the French Quality Order. The IO has established a Quality Assurance Program for the construction of the ITER machine to meet the requirements of the Order and to ensure that ITER activities are performed to achieve the safety and performance objectives of the ITER machine. The requirements in the program shall be followed by all performers involved in the project not only the IO, but DAs and their suppliers and subcontractors. This paper represents the quality requirements from the Order, and roles and responsibilities between each performer involved in the project. The paper also shows the main characteristics of the ITER Quality Assurance Program ensuring that all activities performed for the project conform to established and documented requirements.

Commentary by Dr. Valentin Fuster
2010;():605-609. doi:10.1115/ICONE18-29923.

The Republic of Korea has proposed and designed a helium cooled molten lithium (HCML) test blanket module (TBM) to be tested in the ITER. Ferritic Martensitic Steel (FMS) and beryllium are designed to be used as structural material and armor for the TBM first wall (FW) in this design, respectively. In order to develop the fabrication method for the TBM FW, the joining methods of FMS to FMS and Be to FMS have been developed with hot istotatic pressing (HIP). For joining FMS to FMS, mock-ups were fabricated with an HIP (1050 °C, 100 MPa, 2 hours). For joining Be to FMS, two mock-ups were fabricated with the same method (580 °C, 100 MPa, 2 hours) using different interlayers. Then, in order to evaluate the integrity of the fabricated mock-ups, they were tested at the high heat flux (HHF) test facilities, KoHLTs (Korea Heat Load Test) under 1.0 MW/m2 and 0.5 MW/m2 heat fluxes of up to 1000 cycles.

Topics: Manufacturing
Commentary by Dr. Valentin Fuster
2010;():611-620. doi:10.1115/ICONE18-29943.

In addition to the matured “Laser Inertial Fusion Energy (LIFE)” with spherical compression of deuterium-tritium (DT) for a pure fusion engine or for fusion-fission-hybrid operation, a very new scheme may have now been opened by igniting the neutron-free reaction of proton-boron-11 (p-11 B) using side-on block ignition. Laser pulses of several petawatt power and ps duration led to the discovery of an anomaly of interaction, if the prepulses are cut off by a factor 108 (contrast ratio) to avoid relativistic self focusing. In this case the Bobin-Chu conditions of side-on ignition of solid fusion fuel can be applied after several improvements leading to energy gains of 10,000 similar to the Nuckolls-Wood ignition with extremely intense 5 MeV electron beams. In contrast to the impossible laser-ignition of p-11 B by the usual spherical compression, the side-on ignition is less than ten times only more difficult of DT ignition. This p-11 B fusion produces less radioactivity per gained energy than burning coal. After encouraging success with computations based on the different nuclear cross sections, next steps are focusing on stability and transport problems.

Commentary by Dr. Valentin Fuster
2010;():621-627. doi:10.1115/ICONE18-29945.

Following the first result of generating nuclear fusion energy without dangerous radioactive radiation by laser ignition of the proton-11 Boron reaction (HB11), we applied this method to evaluate other fusion reactions with no primary neutron production as the proton-7 Lithium reaction (HLi7) and of the burning of solid density helium isotope 3 He (He3-He3). The new method is a combination of now available laser pulses of 10 petawatt (PW) power and duration in the range of picoseconds (ps) or less. The new mechanism follows the initial theory of Chu and of Bobin for side-on ignition of solid state density fusion fuel developed in about 1972 where some later known physics phenomena had to be added. The essential innovation is the use of the discovery of a predicted anomaly when the mentioned laser pulses are sufficiently clean, i.e. free from prepulses by at least a contrast ratio 108 where acceleration by the nonlinear (ponderomotive) force is dominating.

Commentary by Dr. Valentin Fuster
2010;():629-634. doi:10.1115/ICONE18-29959.

Fast Ignition is recognized as the most promising approach to achieving the high energy gain target performance needed for commercial inertial confinement fusion (ICF). However, there are great difficulties related to the traditional approach of generating a relativistic electron beam and then focusing it on the hot “spark” region (hot spot) of the compressed target. One promising alternate approach that has been proposed by researchers at LLNL and LANL is the laser generation of a proton beam in an “interaction foil” to ignite the target. However, the total proton flux supplied from hydrogen adsorbed on the foil surface is too small to generate the desired proton flux threshold. In more recent studies it has been found that ions heavier than protons would provide better focusing on the hot spot if ways to efficiently generate them can be found. Here we propose to utilize a new “Deuterium Cluster ”type structure as the laser interaction foil to generate an energetic deuteron beam as the fast igniter. The ultra high density deuterium in the cluster structure promises much higher total flux for deuterons than for traditional protons. Also, deuterons will serve very important dual purposes — the deuteron deposition in the target hot spot will not only provide heating but also fuse with fuel as they slow down in the target. The resulting fusion alphas serve to provide added heating, reducing the input requirement. If the physics works as anticipated, this novel type of interaction foil can efficiently generate energetic deuterons during intense laser pulses. The massive yield of deuterons generated from our cluster material through laser acceleration, should turn out to be the most efficient way of FI of the DT fuel, and making the dream of near-term commercialization of FI fusion more achievable.

Topics: Deuterons , Ignition
Commentary by Dr. Valentin Fuster
2010;():635-639. doi:10.1115/ICONE18-30032.

Today, the tokamak has emerged as the leading approach to controlling nuclear fusion for the purpose of electrical power generation. As an important power system in experimental advanced superconducting tokamak (EAST), an advanced steady-state plasma physics experimental device, the lower hybrid current drive (LHCD) system provides a high-energy microwave for plasma heating and current drive. The microwave power is delivered to the plasma through the lower hybrid wave (LHW) antenna. In order to couple microwave energy to the plasma more efficiently, the antenna is exposed to the plasma. During the plasma operation, the LHW antenna will not only withstand the high thermal flux from the plasma but also the thermal stress owing to thermal loads. The temperature of the antenna has been analyzed using the numerical analysis method. At the same time, the thermal stresses and displacements due to thermal loads are also calculated using the finite element code. The paper is organized as follows: First, the structure of the LHW antenna is briefly described; the 3D model is given. Secondly, thermal loads and boundary conditions are shown and discussed. Finally, the temperature, thermal stresses and displacements are simulated. All the results of the simulation are presented and discussed. These could be useful for the development of the lower hybrid wave antenna.

Topics: Tokamaks , Waves
Commentary by Dr. Valentin Fuster
2010;():641-644. doi:10.1115/ICONE18-30048.

A conceptual design on neutronics of Fusion-Fission Hybrid Reactor (FFHR) blanket is presented in this paper. The code system COUPLE2.0, which couples the codes MCNP and ORIGEN2 is uesd to simulate the depletion of nucleus for FFHR blanket. By comparing different kinds of coolant, fission fuel and tritium breeding material etc, light water was selected as the coolant, the U-Zr alloy as fuel and Li4SiO4 as tritium breeding material. The volumetric ratio of showed that under that the coolant to fuel is approximately 2.0. The tritium breeding ratio (TBR) must be larger than 1.0, the keff and the energy multiplication factor of the blanket are both comparatively high, so that high breeding ratio of fuel can obtained under realistic condition of fusion plasma parameters.

Commentary by Dr. Valentin Fuster
2010;():645-649. doi:10.1115/ICONE18-30143.

This paper overviews the R&D activity of Water Cooled Ceramic Breeder (WCCB) Blanket in Japan. Japan is performing R&D of WCCB Blanket as the primary candidate of the breeding blanket for the fusion DEMO reactor. Regarding the development of blanket module fabrication technology, a real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS) F82H. Using fabricated FW mockup, thermo-hydraulic performance and high heat flux tests were successfully performed with the heat flux equivalent to ITER TBM condition, 0.5 MW/m2 , 80 cycles with the coolant condition as DEMO, 15 MPa 300 °C. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure have been successfully fabricated. Furthermore, assembling of the real scale FW plate mockup and SW plate mockup was successfully performed. Development of major key technologies for the WCCB TBM structure fabrication has been progress toward the establishment of fabrication technology of WCCB TBM.

Commentary by Dr. Valentin Fuster
2010;():651-657. doi:10.1115/ICONE18-30174.

Mini-TBM will be tested in chinese LiPb experimenttal loop Dragon-IV to validate the thermal-hydraulic effect of DFLL-TBM, such as dual-flow fields heat transfer, temperature fields, velocity fields, flux distribution of liquid lithium lead and helium gas. It is difficult to measure the detailed dual-flow fields of liquid metal LiPb and helium gas in mini-TBM. Three dimensions numerical analysis of the LiPb and helium gas flow and heat transfer in Mini-Test Blanket Module (TBM) therefore has been carried out using the CFD code FLUENT. The detailed dual-flow fields, which include temperature, velocity, pressure and heat transfer of liquid LiPb and helium gas, are presented to support for the test of mini-TBM, and to supply more robust database and make a significant joint contribution to the future TBM testing in EAST and ITER, and also optimize and improve the design of DFLL-TBM system for ITER.

Commentary by Dr. Valentin Fuster
2010;():659-665. doi:10.1115/ICONE18-30368.

The development of a unique long cylindrical neutron source for broad area neutron activation analysis (NAA) is presented. This source uses inertial electrostatic confinement (IEC) to produce 2.54 MeV D-D or 14.1 MeV D-T fusion neutrons for applications ranging from security inspection stations to driven-subcritical research assemblies. This design uses a biased grid to initial in a unique “star” mode plasma discharge forming beam-background gas (target) fusion. In spherical geometry it routinely produces ∼108 2.54-MeV D-D fusion neutrons/s at steady-state. Pulsed operation has achieved up to 109 neutrons/sec. (equivalent to 1011 n/s using D-T fill). Indeed, a version of the spherical IEC has been produced commercially as a portable neutron source for industrial NAA applications. Recently a cylindrical (2-dimensional version) design based on the spherical unit has been developed. This provides a unique long “line-like” neutron source for use in broad area NAA. This IEC forms ion beams in the volume between the grounded wall and the concentric cylindrical grid. Those beams converge in the center, much like in the star mode spherical IEC. To date, neutron yields of up to 108 D-D neutrons/sec have been achieved with the cylindrical device. A sealed-off unit using getters for gas storage-control has been developed to simplify use in practical applications such as a luggage inspection station. Such units would be filled with deuterium at a central fueling facility, and sent out to the field. After extended operation, they would be returned to this facility for refilling.

Commentary by Dr. Valentin Fuster

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