Experimental Studies on Pb-Bi Flow Technology and Steel Compatibility With Pb-Bi

For the development of the Pb-Bi cooled FBR and the ADS with Pb-Bi target, the compatibility of steels for core, structural and window materials with high temperature Pb-Bi is one of the critical issues. The effect of corrosion on the mechanical strength of steels should be also concerned. An oxide layer which is formed and self-healed on the steel surface in Pb-Bi is expected to improve the compatibility if oxide potential in Pb-Bi is controlled and monitored adequately to form stable oxide layer. Therefore, monitoring technology of oxygen concentration in Pb-Bi is required. In the present study, a performance test of oxygen sensor, a steel corrosion test and a steel mechanical strength test, or a pipe rupture test, were performed as follows: (1) Test of oxygen sensor: For the monitor of the oxygen potential in Pb-Bi, a thermal stress proof type oxygen sensor made of electrolyte conductor (MgO-ZrO2 and Y2O3-ZrO2) with the reference fluid of oxygen saturated bismuth was developed, and the performance test was conducted using the corrosion test loop. The performance was stable and reliable in the 1000-hour operation. The electromotive forces (EMF) of the sensor cells of MgO-ZrO2 and Y2O3-ZrO2 were nearly the same as each other, and they were not destructed during the 1000-operation. (2) Steel corrosion test: High Cr steels including heat resisting steels were exposed to a liquid Pb-Bi flow at the temperature of 550°C, the velocity of 1m/s, the oxygen concentration of 1.7×10−8wt% and the temperature difference of 150°C for 1000 and 500 hours. It was found that weight losses were lower in general in the steels with higher Cr content. The steels with high Cr, Si and Al formed thin oxide layers and exhibited better compatibility with Pb-Bi. (3) Steel mechanical strength test (pipe rupture test): Metallurgical analysis for ruptured pipe made of SS-316 was performed. The pipe had experienced the exposure to Pb-Bi at 400°C for 3440 hours, at 350°C for 4 hours, at 300°C for 50 hours, and at 250°C for 622 hours. Pipe rupture occurred possibly due to thermal expansion of Pb-Bi at heat-up processes. The results of the analysis indicated that Pb-Bi penetration to steel matrix occurred more seriously near the ruptured part than the other part of the pipe. The analytical result suggested that a brittle fracture might occur in the inner part of the ruptured pipe wall by liquid metal embrittlement because of Pb-Bi penetration, whereas dimples observed suggested that ductile fracture might occur in the outer part of the ruptured pipe wall.

The Design Features of Integrated Modular Water Reactor (IMR)

Small-to-medium-sized (300–600MWe) reactors are required for the electric power market in the near future (2010–2030). The main theme in the development of small-to-medium-sized reactor is how to realize competitive cost against other energy sources. As measures to this disadvantage, greatly simplified and downsized design is needed. From such point of view, Integrated Modular Water Reactor (IMR), which electric output power is 350 MWe, adopts integrated and high temperature two-phase natural circulation system for the primary system. In this design, main coolant pipes, a pressurizer, and reactor coolant pumps are not needed, and the sizes of a reactor vessel and steam generators are minimized. Additionally, to enhance the economy of the whole plant, fluid system, and Instrumentation & Control system of IMR have also been reviewed to make them simplest and smallest taking the advantage of the IMR concept and the state of the art technologies. For example, the integrated primary system and the stand-alone direct heat removal system make the safety system very simple, i.e., no injection, no containment spray, no emergency AC power, etc. The chemical and volume control system is also simplified by eliminating the boron control system and the seal water system of reactor coolant pumps. In this paper, the status of the IMR development and the outline of the IMR design efforts to achieve the simplest and smallest plant are presented.

CAREM X INPRO Case Study

Argentina is performing CAREM X Nuclear System Case Study based on CAREM nuclear reactor and Once Through Fuel Cycle, using SIGMA for enriched uranium production, and a deep geological repository for final disposal of high level waste after surface intermediate storage in horizontal natural convection silos, to verify INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) methodology. Projections show that developing countries could play a crucial role in the deployment of nuclear energy, in the next fifty years. This case study will be highly useful for checking INPRO methodology for this scenario. In this contribution to ICONE 12, the preliminary findings of the Case Study are presented, including proposals to improve the INPRO methodology.

A Review of Root Causes of SCC Phenomena in BWR/RBMK: Issues, Approaches, and Proposals

The author looked for potential mechanisms deemed insignificant in the early stages of SCC research and the current approach used in SCC studies were selected. The basis for those mechanisms was in non-nuclear corrosive environments as well as analyses of redox potential data measured under radiation. Three possible mechanisms were identified; ‘long cell action (which suppresses local cell cathodic activities and accelerates remote local cell anodic activities),’ ‘autocatalytic growth of cracks in crevice water chemistry,’ and ‘transpassive corrosion of stainless steel.’ The ‘long cell action’ has been identified as a key mechanism of soil corrosion which is induced by a difference in the availability of oxygen inside the soil. In this mechanism, conduction of electrons through an electric conductor over a long distance plays a significant role. This author proposes a hypothetical mechanism that states; ‘radiation-induced ‘long cell action’ causing electrolytic corrosion.’

Protective Insulated Coating for IGSCC Mitigation in BWRs

A protective insulated coating (PIC) on 304 stainless steel (SS) surfaces as an IGSCC mitigation method was developed and investigated in high temperature water under various water chemistry conditions by measuring the electrochemical corrosion potential (ECP) and flow-assisted corrosion (FAC) rate. The ECP results clearly demonstrate that the PIC layer restricted oxidant transport to the metal surface, and the ECP remained at <−230 mV (SHE) in 288°C containing high oxygen (O2) and no hydrogen (H2). In this paper, long term durability of PIC layer prepared by various coating methods will be discussed.

Modelling and Dynamics Analysis of Heat Exchanger as a Distributed Parameter Process

On the basis accepted and critically clarified assumptions, a non–linear and afterwards linearized mathematical model of fuel oil cooling chamber has been developed in engineering sense sufficiently correct. The model is in the form of set of partial differential equations with constant coefficients. Using the appropriate numerical simulation of the results derived, the dynamic of this process has been shown in the form of appropriate transient processes responses which quite well correspond to the real process behavior.

Hydrogen Generation Using the Modular Helium Reactor

Process heat from a high-temperature nuclear reactor can be used to drive a set of chemical reactions, with the net result of splitting water into hydrogen and oxygen. For example, process heat at temperatures in the range 850°C to 950°C can drive the sulfur-iodine (SI) thermochemical process to produce hydrogen with high efficiency. Electricity can also be used to split water, using conventional, low-temperature electrolysis (LTE). An example of a hybrid process is high-temperature electrolysis (HTE), in which process heat is used to generate steam, which is then supplied to an electrolyzer to generate hydrogen. In this paper we investigate the coupling of the Modular Helium Reactor (MHR) to the SI process and HTE. These concepts are referred to as the H2-MHR. Optimization of the MHR core design to produce higher coolant outlet temperatures is also discussed.

User Requirements and Criteria for Proliferation Resistance in INPRO

In designing future nuclear energy systems, it is important to consider the potential that such systems could be misused for the purpose of producing nuclear weapons. INPRO set out to provide guidance on incorporating proliferation resistance into innovative nuclear energy systems (INS). Generally two types of proliferation resistance measures are distinguished: intrinsic and extrinsic. Intrinsic features consist of technical design features that reduce the attractiveness of nuclear material for nuclear weapon program, or prevent the diversion of nuclear material or production of undeclared nuclear material for nuclear weapons. Extrinsic measures include commitments, obligations and policies of states such as the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) and IAEA safeguards agreements. INPRO has produced five basic principles and five user requirements for INS. It emphasizes that INS must continue to be an unattractive means to acquire fissile material for a nuclear weapon program. It also addresses as user requirements: 1) a balanced and optimised combination of intrinsic features and extrinsic measures, 2) the development and implementation of intrinsic features, 3) an early consideration of proliferation resistance in the development of INS and 4) the utilization of intrinsic features to increase the efficiency of extrinsic measures. INPRO has also developed criteria, consisting of indicators and acceptance limits, which would be used by a state to assess how an INS satisfies those user requirements. For the first user requirement, the most important but complex one, INPRO provides a 3-layer hierarchy of indicators to assess how unattractive a specific INS would be as part of a nuclear weapon program. Attributes of nuclear material and facilities are used as indicators to assess intrinsic features. Extrinsic measures imposed on the system are also assessed. Indicators to assess defence in depth for proliferation resistance include the number and robustness of barriers, and the redundancy or complementarity of barriers. The cost of incorporating proliferation resistant features is used to assess the cost-effectiveness of any particular INS in providing proliferation resistance. The stages in the development of an INS at which proliferation resistance is considered in the process are assessed. Awareness of extrinsic measures by designers and use of intrinsic features for verification illustrate how intrinsic features facilitate extrinsic measures. An INPRO-consistent methodology to assess the proliferation resistance of an INS is still under development, with feedback expected from the case studies undertaken by Argentina, India, Russia and the Republic of Korea.

Results of INPRO in the Area of Economics

In the area of Economics four selected scenarios from the SRES study have been analysed within the International Project on Innovative Reactors and Fuel Cycles (INPRO) of the IAEA. They cover a range of possible future developments characterized by different degrees of globalisation and by different relative priorities on economic and environmental objectives. Four “aggressive nuclear” variants, one for each of the four selected SRES scenarios, are also analyzed. Provided innovative nuclear energy systems (INS) are economically competitive, they can play a major role in meeting future energy needs. Future economic competitiveness will depend on the speed of continuing cost reductions achieved by nuclear energy relative to competing technologies. The paper presents specific capital costs and electricity production costs at which nuclear energy is competitive in 2050 in the four selected SRES scenarios, and estimates corresponding costs for nuclear energy in the four aggressive nuclear variants. The important message is that for nuclear technology to gain and grow market share it must benefit sufficiently from learning to keep it competitive with competing energy technologies. For such learning to take place experience must be gained and to gain such experience the energy from INS must be cost competitive with energy from alternative sources and INS must represent an attractive investment to compete successfully in the capital market place. In total, INPRO defined two basic principles, five user requirements and several criteria in this area, which are presented in the full paper. To be cost competitive all component costs, e.g., capital costs, operating and maintenance costs, fuel costs, must be considered and managed to keep the total unit energy cost competitive. Limits on fuel costs in turn imply limits on the capital and operating cost of fuel cycle facilities, including mines, fuel processing and enrichment, fuel reprocessing and the decommissioning and long term management of the wastes from these facilities. Cost competitiveness of energy from INS will contribute to investor confidence, i.e. to the attractiveness of investing in INS, as will a competitive rate of return.

Projection of the Neutronic and Thermal Fuel Rod Behavior in a BWR

Currently, at the Instituto Nacional de Investigaciones Nucleares (National Institute for Nuclear Research) in Mexico, it is being developed a computational code for evaluating the neutronic, thermal and mechanical performance of a fuel element at several different operation conditions. The code is referred as to MCTP (Multigrupos con Temperaturas y Potencia), and is benchmarked against data from the Laguna Verde Nuclear Power Plant (LVNPP). In the code, the neutron flux is approximated by six groups of energy: one group in the thermal region (E < 0.625 eV), four in the resonances region (0.625 eV < E < 0.861 MeV), and one group in the fast region (E > 0.861 MeV). Thus, the code is able to determine the damage to the cladding due to fast neutrons. The temperature distribution is approximated in both axial and radial directions taking into account the changes in the coolant density, for both the single and two-phase regions in a BWR channel. It also considerate the changes in the thermal conductivity of all materials involved for the temperature calculations, as well as the temperature and density effects in the neutron cross sections. In the code, fuel rod burnup is evaluated. Also, plutonium production and poison production from fission. In this work, the neutronic and thermal performance of fuel rods in a 10×10 fuel assembly is evaluated. The fuel elements have a content of 235U. The fuel assembly was introduced to the unit 1 of LVNPP reactor core in the cycle 9 of operation, and will stay in during three cycles. In the analysis of fuel rod performance, the operating conditions are those for the cycle 9 and 10, whereas for the current cycle (cycle 11) the reactor is projected to operate during 460 days. The analysis for cycle 11 uses the actual location of the fuel assembly that will have in the core. The results show that the fuel rods analyzed did not reach the thermal limits during the cycles 9 and 10, as expected, and for cycle 11 the same thermal limits are not predicted to be reached.