Long life multipurpose small size fast reactor with liquid metallic-fueled core

A concept of a long life multipurpose nuclear reactor with self-sustained liquid metallic fuel is proposed to meet the requirements for the future energy production. The conceptual design is described and the core neutronic characteristics are obtained based on two-dimensional cylindrical diffusion reactor model. The influence of fission products separation in the liquid-fueled system on the core burnup capability is discussed. The burnup analysis shows a feasibility of the long life refueling-free core concept.     

The fuel element of the HTR-module, a prerequisite of an inherently safe reactor

To retain fission products after postulated accidents, power reactors usually rely on active safety systems inside the primary circuit, such as e.g. redundant shut down systems and multiple redundant decay heat removal systems. The HTR-Module is employing a different approach which relies entirely on the ability of the coated particle to retain all key radio-nuclides as long as a certain maximum fuel element temperature is not exceeded. Consequently, the reactor is designed such that for any postulated accident this maximum fuel element temperature is not reached even without relying on any active safety systems inside the primary circuit, since the decay heat can be removed to an outside heat sink solely by passive means.The paper discusses the experimental results of fission product release from spherical fuel elements for various temperatures. From the tests as well as from statistical considerations it can be concluded that any hazardous radiation dose to the environment can be excluded if the maximum fuel element temperature in the HTR-Module stays below 1600°C.     

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ_ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10³Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.     

A design study on MOX-fueled small fast reactors for standardization of a small fast nuclear reactor system

A way of development to standardize a small fast nuclear reactor system, which is considered one of the suitable concepts at next generation for satisfying such needs as generality, small dependence on natural resources, safety and non-proliferation, is proposed. This process consists of three steps : the first is to demonstrate the basic system within a short period based on current techniques, the second is to achieve greatly higher economy, and the final is to standardize the commercial system that can economically compete with or overcome current light water reactors. A technical investigation is conducted on the performance of a mixed-oxide (MOX)-fueled small fast reactor with a reflector-driven reactivity control system to satisfy the needs at the first step, considering plenty of accomplishments on the MOX fuel and its advantage for limiting the duration of development to the level required at the stage. The results obtained from a series of neutronic and thermal-hydraulic calculations show the feasibility of a small fast reactor that produces the electric power of about 50MW, achieves about two-year consecutive operation with high safety performance and is greatly flexible for updating the system. A mixed-nitride-fueled core is found to be promising past the first stage.     

The inherently-safe power reactor DYONISOS : Dynamic nuclear inherently-safe reactor operating with spheres

A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydrodynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h.     

The advanced high-temperature reactor: High-temperature fuel, liquid salt coolant, liquid-metal-reactor plant

The Advanced High-Temperature Reactor is a new reactor concept that combines four existing technologies in a new way: (1) coated-particle graphite-matrix nuclear fuels (traditionally used for helium-cooled reactors), (2) Brayton power cycles, (3) passive safety systems and plant designs from liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants with boiling points far above the maximum coolant temperature. The new combination of technologies enables the design of a large [2400- to 4000-MW(t)] high-temperature reactor, with reactor-coolant exit temperatures between 700 and 1000°C (depending upon goals) and passive safety systems for economic production of electricity or hydrogen. The AHTR [2400-MW(t)] capital costs have been estimated to be 49 to 61% per kilowatt (electric) relative to modular gas-cooled [600-MW(t)] and modular liquid-metal-cooled reactors [1000-MW(t)], assuming a single AHTR and multiple modular units with the same total electrical output. Because of the similar fuel, core design, and power cycles, about 70% of the required research is shared with that for high-temperature gas-cooled reactors.     

The role of the reactor size for an investment in the nuclear sector: An evaluation of not-financial parameters

The literature presents many studies about the economics of new Nuclear Power Plants (NPPs). Such studies are based on Discounted Cash Flow (DCF) methods encompassing the accounts related to Construction, Operation & Maintenance, Fuel and Decommissioning. However the investment evaluation of a nuclear reactor should also include not-financial factors such as siting and grid constraints, impact on the national industrial system, etc.The Integrated model for the Competitiveness Assessment of SMRs (INCAS), developed by Politecnico di Milano cooperating with the IAEA, is designed to analyze the choice of the better Nuclear Power Plant size as a multidimensional problem. In particular the INCAS’s module “External Factors” evaluates the impact of the factors that are not considered in the traditional DCF methods.This paper presents a list of these factors, providing, for each one, the rationale and the quantification procedure; then each factor is quantified for the Italian case. The IRIS reactor has been chosen as SMR representative.The approach and the framework of the model can be applied to worldwide countries while the specific results apply to most of the European countries. The results show that SMRs have better performances than LRs with respect to the external factors, in general and in the Italian scenario in particular.     

The Canadian approach to safe, permanent disposal of nuclear fuel waste

The assessment of nuclear fuel waste disposal deep in plutonic rock of the Canadian Precambrian Shield is now well advanced. A comprehensive understanding has been developed of the chemical and physical processes controlling the containment of radionuclides in used fuel. The following conclusions have been reached:

• - Containers with outer shells of titanium or copper can be expected to isolate used fuel from contact with groundwater for at least 500 years, the period during which the hazard is greatest.

• - Uranium oxide fuel can be expected to dissolve at a rate less than 10⁻⁸ per day, resulting in very low rates of radionuclide release. This is consistent with observations of uranium oxide deposits in the earth's crust.

• - Transport of radionuclides away from the containers can be significantly delayed by placing a compacted bentonite-clay based layer between the container and the rock mass.

• - The granite plutons of interest consist of relatively large rock volumes of low permeability separated by relatively thin fracture zones. The low permeability volumes are sufficiently large to accommodate a vault design that will ensure radionuclides do not reach the surface in unacceptable concentrations.

Our field and laboratory investigations, together with assessments of conceptual disposal vault designs, give us confidence that the combination of engineered barriers and a technically suitable plutonic rock site will meet the requirements for safe disposal of nuclear fuel wastes in Canada.     

Safety assessment for computerized nuclear reactor protection systems: The Markov approach

In reactor protection systems based on minicomputers a central role is played by the diagnostic capability of selfchecking programs. It is thus of great importance to determine the efficiency that such programs must have with respect to fault detection in order to meet a certain reliability goal. Even though the content of this report is part of the safety study on a particular plant (Tapiro Research Reactor in service at C.S.N. Casaccia) it allows one to reach more general conclusions about the reliability of computerized protection systems. Another major aim of this paper is to point out the methodological difficulties met in the safety qualification of these systems.     

Overview of plant specific severe accident management strategies for Kozloduy nuclear power plant,WWER-1000/320

This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.     

The Nuclear Battery: a very small reactor power supply for remote locations

The Nuclear Battery is a small reactor power supply being developed by Atomic Energy of Canada Limited for use in locations that are remote from utility grids and natural gas pipelines. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low enriched uranium TRISO coated-particle fuel and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy that may be converted into electricity in an organic Rankine cycle engine, or used to produce high-pressure steam. The reference design is capable of producing about 2400 kW(t) (about 600 kW(e) net) for 15 full-power years without refuelling.     

A heat, mass, and momentum transport model for hydrogen diffusion flames in nuclear reactor containments

It is now possible to analyze the time-dependent, fully three-dimensional behavior of hydrogen combustion in nuclear reactor containments. This analysis involves coupling the full Navier-Stokes equations with multi-species transport to the global chemical kinetics of hydrogen combustion. A transport equation for the subgrid scale turbulent kinetic energy density is solved to produce the time and space dependent turbulent transport coefficients. The heat transfer coefficient governing the exchange of heat between fluid computational cells adjacent to wall cells is calculated by a modified Reynolds analogy formulation. The analysis of a MARK-III containment indicates very complex flow patterns that greatly influence fluid and wall temperatures and heat fluxes.     

The effect of inspection and repair on the size distributions of the weld imperfections in nuclear power plant pipes

The weld defects appearing in the pipes of the main steam system in a BWR power plant were studied. The effect of inspection rejection and repair on the imperfection size distribution was analysed. The size distributions before and after the inspection rejection as well as after the repair procedure were of the form a + b/x², where x is the imperfection size and a and b are coefficients. The portion of the rejected defects had the size dependence near to the form of the cumulative Gaussian function. The effect of repair on the initial size distribution had the form of the cumulative Poisson distribution.     

The determination of plant status during abnormal reactor operating conditions: Accident sequence identification

A study was performed to determine information available to reactor operators for assessing core status during severe accidents. Beginning with a simplified accident event tree describing possible final plant states, an instrumentation availability matrix was developed to identify potential gaps in current nuclear plant instrumentation. This matrix was further developed for a representative PWR system. The major study conclusion is that the largest potential gap lies in methods available for directly determining the stage of core degradation. This inadequacy influences the potential effectiveness of off-site emergency response planning, and also the ability of plant personnel to mitigate against the further worsening of the plant condition. Possible methods for improving direct indications of core damage are discussed with the recommendation that further study and experimentation be conducted in this area.     

NURESIM - A European simulation platform for nuclear reactor safety: Multi-scale and multi-physics calculations, sensitivity and uncertainty analysis

Starting in 2005 with the NURESIM Integrated Project (FP6), a European Reference Simulation Platform for Nuclear Reactors called NURESIM is being developed. This development follows a roadmap which is consistent with the SRA (Strategic Research Agenda) of the European SNETP (Sustainable Nuclear Energy Technology Platform). After delivery of two successive versions during the course of the NURESIM project, the numerical simulation platform is presently being developed in the frame of the NURISP European Collaborative Project (FP7), which includes 22 organizations from 14 European countries.NURESIM intends to be a reference platform providing high quality software tools, physical models, generic functions and assessment results.The NURESIM platform provides an accurate representation of the physical phenomena by promoting and incorporating the latest advances in core physics, two-phase thermal-hydraulics and fuel modelling. It includes multi-scale and multi-physics features, especially for coupling core physics and thermal-hydraulics models for reactor safety. Easy coupling of the different codes and solvers is provided through the use of a common data structure and generic functions (e.g., for interpolation between nonconforming meshes).More generally, the platform includes generic pre-processing, post-processing and supervision functions through the open-source SALOME software, in order to make the codes more user-friendly.The platform also provides the informatics environment for testing and comparing different codes. For this purpose, it is essential to permit connection of the codes in a standardized way. The standards are being progressively built, concurrently with the process of developing the platform.The NURESIM platform and the individual models, solvers and codes are being validated through challenging applications corresponding to nuclear reactor situations, and including reference calculations, experiments and plant data. Quantitative deterministic and statistical sensitivity and uncertainty analyses tools are also developed and provided through the platform.A Users’ Group of European and non-European countries, including vendors, utilities, TSOs, and additional research organizations (beyond the current partners) has also been established in order to enhance the role of the simulation platform in meeting the needs of the nuclear industry, as applied to current and future nuclear reactors.This presentation summarizes the achievements and ongoing developments of the simulation platform in core physics, thermal-hydraulics, multi-physics, uncertainties and code integration.     

Materials-technology investigations of fuel elements, irradiated in a reactor at the obninsk nuclear power plant, after standing for 38 years in a depository

The results of materials-technology investigations of a spent fuel assembly from a reactor at the Obninsk nuclear power plant, the first nuclear power plant in the world, before the rated burnup and after prolonged dry storage (for about 40 years) were presented. It was established that the fuel elements from the fuel assembly studied are in satisfactory condition. No appreciable damage due to the prolonged storage was found: the outer diameter remains within the technological tolerance limits and the strength and the plasticity of the jackets are high. Only surface corrosion damage to 10 μm depth was found on the fuel-element jackets. The fuel composition remained whole. 6 figures, 1 table, 3 references. State Science Center of the Russian Federation—A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 88, No. 3, pp. 183–188, March, 2000.     

The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

Environmentally assisted cracking (EAC) or, in other words, stress corrosion cracking (SCC) of in-core materials has become an increasingly important reason for the downtime and maintenance costs of nuclear power plants (NPPs). Use of small size specimens for stress corrosion testing of irradiated materials is necessary because handling of high dose rate materials is difficult and the availability of these materials is limited. A drawback of using small size specimens is that they do not in some cases fulfil the requirements of the relevant testing standards and sometimes their limited load-bearing capacity prevents corrosion fatigue tests and tests with static loading at reasonable K_I values. The test results show that the ductile fracture resistance curves of a Cu–Zr–Cr alloy are, to some extent, independent of the specimen geometry and size. However, the curves of small specimens deviate from the curves of larger specimens at high J values (large plastic zone relative to the remaining ligament) or when the crack growth exceeds about 30% of the remaining ligament. The size dependency of the tested Cu–Zr–Cr alloy seems to be a consequence of decreasing stress triaxiality as the size of the specimen is decreased. The results of the SCC tests of sensitized SIS 2333 stainless steel (equal to AISI 304) specimens in simulated boiling water reactor (BWR) water show that the plastic deformation of the remaining ligament of the specimen has no significant effect on the environmentally assisted crack growth rate. This indicates that stress corrosion testing is not limited by the specimen size. The size dependency in SCC tests should be further studied by conducting tests using various specimen sizes.