Homogenisation method for the modal analysis of a nuclear reactor with internal structures modelling and fluid–structure-interaction coupling

A homogenisation method is presented and validated in order to perform the dynamic analysis of a nuclear pressure vessel with a “reduced” numerical model accounting for inertial fluid–structure coupling and describing the geometrical details of the internal structures, periodically embedded within the nuclear reactor. Homogenisation techniques have been widely used in nuclear engineering to model confinement effects in reactor cores or tubes bundles. Application of such techniques to rector internals is investigated in the present paper. The theory bases of the method are first recalled. Adaptation of the homogenisation approach to the case of rector internals is then exposed: it is shown that in such case, confinement effects can de modelled by a suitable modification of classical fluid–structure symmetric formulation. The method is then validated by comparison of 3D and 2D calculations. In the latter, a “reduced” model with homogenised fluid is used, whereas in the former, a full finite element model of the nuclear pressure vessel with internal structures is elaborated. The homogenisation approach is proved to be efficient from the numerical point of view and accurate from the physical point of view. Confinement effects in the industrial case can then be highlighted.     

Prospects for fission gas extraction from in-reactor UO2 nuclear fuel elements

Due to the many problems encountered in the design of fuel rods for the safe operation of commercial nuclear reactors, caused by the fission gases generated by the fission of fissile material, it was considered opportune to make a theoretical analysis of the feasibility of extraction of fission gases from the fuel rod while in operation.This analysis in the steady state of a Zircaloy-2 sheathed fuel rod containing UO₂ as a fuel, with a 2 mm (2.7 vol.%) diameter porous graphite cylinder inserted in the centre, has demonstrated that a total volume of fission gases (xenon, krypton, and iodine) of about 1.1 × 10⁻⁶ cm³/s (at STP) can be extracted from the fuel rod at a controlled rate, determined by the inherent property of fission gas migration towards the centre of the fuel rod from its place of formation. In this analysis, the fuel rod was assumed to be subjected to irradiation in a reactor the size of a Bruce “A” reactor, operating at 3000 megawatts thermal power. The extracted volume of gas was calculated on a 900 h cycle after the first 90 h of reactor operation had elapsed.     

Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation in an extended operation domain with increased void and thereby increased void reactivity feedback and which often have thinner fuel rods and thereby decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of “unexpected” instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, the subject of BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a “new and improved” state of the art has emerged recently.     

Risk-benefit evaluation of nuclear power plant siting

An assessment scheme is described for the risk-benefit analyses of nuclear power versus conventional alternatives. Given the siting parameters for the proposed nuclear plant an economic comparison is made with the most advantageous competitive conventional production scenario. The economic benefit is determined from the differential discounted annual energy procurement cost as a function of the real interest rate and amortization time. The risk analysis encompasses following factors: radiation risks in normal operation, reactor accident hazards and economic risks, atmospheric pollutants from the conventional power plants and fuel transportation. The hazards are first considered in terms of probabilistic dose distributions. In the second stage risk components are converted to a compatible form where excess mortality is used as the risk indicator. Practical calculations are performed for the power production alternatives of Helsinki where district heat would be extracted from the nuclear power plant. At the real interest rate of 10% and amortization time of 20 yr the 1000 MW(e) nuclear option is found to be $9.1 m per yr more economic than the optimal conventional scenario. Simultaneously the nuclear alternative is estimated to reduce excess mortality by 2–5 fatal injuries annually.     

Computational Fluid Dynamics for nuclear applications: from CFD to multi-scale CMFD

New trends in computational methods for nuclear reactor thermal–hydraulics are discussed; traditionally, these have been based on the two-fluid model. Although CFD computations for single phase flows are commonplace, Computational Multi-Fluid Dynamics (CMFD) is still under development. One-fluid methods coupled with interface tracking techniques provide interesting opportunities and enlarge the scope of problems that can be solved. For certain problems, one may have to conduct “cascades” of computations at increasingly finer scales to resolve all issues. The case study of condensation of steam/air mixtures injected from a downward-facing vent into a pool of water and a proposed CMFD initiative to numerically model Critical Heat Flux (CHF) illustrate such cascades. For the venting problem, a variety of tools are used: a system code for system behaviour; an interface-tracking method (Volume of Fluid, VOF) to examine the behaviour of large bubbles; direct-contact condensation can be treated either by Direct Numerical Simulation (DNS) or by analytical methods.     

The encapsulated nuclear heat source reactor for low-waste proliferation-resistant nuclear energy

The encapsulated nuclear heat source (ENHS) is a new Pb-Bi cooled modular reactor concept that features a combination of the following useful features that may make nuclear energy more attractive: (1) 20 years of full power operation without refueling. (2) Nearly constant fissile fuel contents and k_eff. (3) No on-site refueling and fueling hardware. (4) The ENHS modules are factory manufactured and transported already fueled to the site. (5) No access to neutrons. (6) No mechanical connections between the ENHS module and the energy conversion plant (The ENHS module has the function of a nuclear battery — with 20 years of full power operation at 125 MW_th). (7) At end of life, the ENHS module serves as a spent fuel storage cask and, later, as a spent fuel shipping cask. That is, the fuel is locked inside the ENHS from “cradle to grave”. (8) 100% natural circulation resulting in passive load following capability and autonomous control. This combination of features offers a highly safe nuclear energy system that is characterized by low waste, high proliferation resistance and high uranium utilization. The low waste and high uranium ore utilization are achieved by recycling the Pu and MA many times using a proliferation-resistant dry process; only fission products are to be extracted between cycles. Spent LWR fuel can provide for the HM make-up. The high level of proliferation resistance is obtained by restricting access to the fuel and neutrons and by eliminating the economic incentive of the client country to invest in sensitive technologies or infrastructure that can be used for clandestine production of strategic nuclear materials.     

Safety consideration and economic advantage of a new underground nuclear power plant design

A conceptual design of an underground nuclear power plant is proposed to make undergrounding of nuclear reactors not only environmentally desirable but also economically feasible. Expedient to the underground environment, this design capitalizes on the pressure-containing and radiation filtering characteristics of the new underground boundary conditions. Design emphasis is on the containment of a catastrophic accident — that of a reactor vessel rupture caused by external means. The igh apacity apid nergy issipation nderground ontainment (HiC—REDUCE) system which efficiently contains loss-of-coolant accidents (LOCAs) and small break conditions is described. The end product is a radiation-release-proof plant which, in effect, divorces the safety of the public from the safety of the reactor.     

A 10 MWe pressurized water reactor for safe and reliable power

A design for an innovative, passively safe 10 MWe power plant based on the proven pressurized water reactor technology has been developed. The plant incorporates an innovative design approach to achieve “walk-away” safety and includes significant simplification and elimination of systems and components when compared to the current generation commercial nuclear power plants. The plant has been designed such that the majority of the equipment will be pre-assembled as modules at off-site facilities and shipped to the site on trucks for installation. This approach will provide shorter construction schedules and improved quality control.     

A simplified refueling scheme for nuclear heating reactor

The nuclear heating reactor is a clean energy resource with high reliability and high economic benefits. Its basic design characteristics are different from those of big reactors. Some features, such as in-pressure vessel spent fuel storage, long refueling period, etc., provide the possibility to simplify the refueling system for economic purposes. After being stored in a pressure vessel for about 15 years, the spent fuel assemblies, with very low radioactivity and decay heat capacity, may be removed from the reactor pressure vessel to a storage pool by a simplified system including a shielded flask, reactor building crane, and some auxiliary tools. It is demonstrated that this ‘dry-method’ refueling scheme is safe and reliable.     

Self-consistent model of nuclear power and nuclear fuel cycle

Under discussion are such major aspects of the nuclear energy sector as cost effectiveness, nuclear and environmental safety of reactors and nuclear fuel cycle facilities, sustained fuel supply, and proven feasibility of a proliferation-resistant technology. These requirements can be met, for instance, by a two-circuit nuclear facility with an inherently safe fast reactor of the BREST type which is expected to produce electricity at a cost not higher than that at modern LWRs. Fuel supply to such facilities and to a relatively small number of thermal reactors with BR<1, could be provided by fast reactors using depleted uranium as makeup fuel and having a small breeding gain in the core (CBR≈1.05) and bottom blanket (full BR≈1.1). Use of a high-boiling metallic coolant (lead) affords deterministic nuclear, technical and environmental safety of the plants in design-basis and hypothetical accidents. Introduction of a transmutational NFC is viewed as one of the avenues to global environmental safety, when the equivalent activity of long-lived high-level waste is made lower or close to the activity of the source material going into energy production. With such a balance in place, nuclear power could be regarded, in a sense, as waste-free.     

Methodologies for rapid evaluation of seismic demand levels in nuclear power plant structures

A methodology for rapid assessment of both acceleration spectral peak and “zero period acceleration” (ZPA) values for virtually any major structure in a nuclear power plant is presented. The methodology is based on spectral peak and ZPA amplification factors, developed from regression analyses of an analytical database. The developed amplification factors are applied to the plant's design ground spectrum to obtain amplified response parameters. A practical application of the methodology is presented.This paper also presents a methodology for calculating acceleration response spectrum curves at any number of desired damping ratios directly from a single known damping ratio spectrum. The methodology presented is particularly useful and directly applicable to older vintage nuclear power plant facilities (i.e. such as those affected by USI A-46). The methodology is based on principles of random vibration theory. The methodology has been implemented in a computer program (SPECGEN). SPECGEN results are compared with results obtained from time history analyses.     

Investigation of the structural behaviour of nuclear spent fuel reprocessing plant components subjected to dynamic loads

In the planned reprocessing plant for spent nuclear fuel elements in Germany components and systems, which form a process-technical unit are integrated in a steel structure called a “module”. In the present paper, the stability and strength of two representative modules under dynamic loading due to earthquake are investigated. Since full scale models of the modules exist, both analytical and experimental investigations are possible. The results of the two investigation methods are in good agreement. Both modules withstand the earthquake loadings with only minor modifications.     

A critical reappraisal of nuclear power plant safety against accidental aircraft impact

The overall problem of nuclear power plant safety against an accidental aircraft impact is discussed in relation with its structural analysis and design. Associated risks, such as fire, which is a potential source of damage for buildings and other structures, are not considered.The paper is divided in two parts. In part I different approaches used for determining the reaction-time curve are discussed. The influence on the results of target motions is examined next. It is shown that for the evaluation of structural response an aircraft-structure interaction analysis is usually an unnecessary refinement, “mean” reaction-time and impact area-time curves being sufficient to define the excitation. Preliminary results for oblique impact are also given. Since the conditional probability of a normal impact is very small, the consideration of oblique impact may become acceptable in future design criteria.In part II, available solutions for the resulting structural dynamic problem are reviewed. The feasibility of resorting to a static analysis is also discussed. Present practices to evaluate floor response spectra are reviewed next. The short-comings of the “deterministic” approach are pointed out. It is proposed to define the excitation as a mean plus a fluctuating force. The latter is treated as a nonstationary random process and the problem solved by numerical integration in the time domain. Although such solutions get prohibitively expensive when the number of degrees of freedom becomes large, results obtained for simple models may help to clarify which are the important variables of the problem.     

Sea-water desalination with nuclear and other energy sources: the EURODESAL project

This paper summarises our recent investigations undertaken as part of the EURODESAL project on nuclear desalination, currently being carried out by a consortium of four European, and one Canadian, industrials and two leading EU R&D organisations.Major achievements of the project, as discussed in this paper are:

• Coherent demonstration of the technical feasibility of nuclear desalination through the elaboration of coupling schemes for optimum cogeneration of electricity and water and by exploring the unique capabilities of the innovative nuclear reactors and desalination technologies.

• Verification that the integrated system design does not adversely affect nuclear reactor safety.

• Development of codes and methods for an objective economic assessment of the competitiveness and sustainability of proposed options through comparison, in European conditions, with fossil energy based systems.

Results obtained so far seem to be quite encouraging as regards the economical viability of nuclear desalination options.Thus, for example, specific desalination costs ($/m³ of desalted water) for nuclear systems, such as the AP-600 and the French PWR-900 (reference base case), coupled to multiple effect distillation (MED) or the reverse osmosis (RO) processes, are 30–60% lower than the desalination costs for fossil energy based systems, using pulverised coal and natural gas with combined cycle, at low discount rates and recommended fossil fuel prices. Even in the most unfavourable scenarios for nuclear energy (discount rate=10%, low fossil fuel costs) desalination costs with the nuclear reactors are 7–20% lower, depending upon the desalination capacities. Furthermore, with the advanced coupling schemes, utilising waste heat from nuclear reactors, the gains in specific desalination costs of nuclear systems are increased by another 2–15%, even without system and design optimisation. A preliminary evaluation shows that desalination costs with the GT-MHR, coupled to a MED process, could still be much lower than the above nuclear options for desalting capacities≤43 000 m³ per day. This is because its design intrinsically provides “virtually free” heat at ideal temperatures for desalination (80–100 °C).     

Hybrid soliton nuclear reactors: A model and simulation (encapsulated long living accelerator driven system)

We intend to explore the potential of Hybrid Soliton Reactors (Réacteur Hybride à Soliton, RHYS) for producing energy. In our case an encapsulated long living fission reactor is driven by a proton accelerator, who produces neutrons on a target. In a first part we give the mathematical approach of such a sub-critical reactor, as an extension of the “Soliton Reactor” which was recently proposed by different authors, Edward Teller, L.P. Feoktistov, and others (H. Sekimoto under the name “Candle reactor”). In a second part we give results of simulations and explore the possibilities to control such a system.     

Development and application of filters for air cleaning in nuclear power plants

The development of filter systems for air cleaning in nuclear power plants will be briefly described. The result of research work on iodine filters was the basis for the use of gasketless deep-bed filters and of the multiway sorption filter in German reactor stations. The composition of the iodine release to the environment was validated with the “discriminating iodine species sampler”. The main sources for the release of elemental iodine from BWR and PWR were discovered and finally suppressed. The mechanical strength of HEPA filters is being tested at high temperatures and humidities. Prototype HEPA filters have been developed with much higher resistance against humidity, differential pressure, and corrosion. A filter for the removal of particles during extreme operating conditions such as containment venting was developed using stainless steel fibers for the filter medium. The first filter of this type has already been built and installed in a modern German PWR as a part of the containment-venting system for serious accidents.     

Deep-Burn: making nuclear waste transmutation practical

In the Deep-Burn concept, destruction of the transuranic component of light water reactor (LWR) waste is carried out in one burn-up cycle, accomplishing the virtually complete destruction of weapons-usable materials (Plutonium-239), and up to 90% of all transuranic waste, including the near totality of Neptunium-237 (the most mobile actinide in the repository environment) and its precursor, Americium-241.Waste destruction would be performed rapidly, without multiple reprocessing of plutonium, thus eliminating the risks of repeated handling of weapons-usable material and limiting the generation of secondary waste. There appears to be no incentive in continuing the destruction of waste beyond this level.An essential feature of the Deep-Burn Transmuter is the use of ceramic-coated fuel particles that provide very strong containment and are highly resistant to irradiation, thereby allowing very extensive destruction levels (“Deep Burn”) in the one pass, using gas-cooled modular helium reactor (MHR) technology developed for high-efficiency energy production. The fixed moderator (graphite) and neutronically transparent coolant (helium) provide a unique neutron energy spectrum to cause Deep-Burn, and inherent safety features, specific to the destruction of nuclear waste, that are not found in any other design.Deep-Burn technology could be available for deployment in a relatively short time, thus contributing effectively to waste problem solutions. Extensive modeling effort has led to conceptual Deep-Burn designs that can achieve high waste destruction levels (70% in critical mode, 90% in with a supplemental subcritical step) within the operational envelope of commercial MHR operation, including long refueling intervals and the highly efficient production of energy (approximately 50%). To the plant operator, a Deep-Burn Transmuter will be identical to its commercial reactor counterpart.     

Stability of a nuclear power generation plant with circulating fuel

A nuclear power plant with circulating incompressible fuel is discussed. The active zone of the reactor is idealized by a system with lumped parameters and the heat exchanger by a unit with distributed parameters. The stability of such a plant over a small region of the steady state cycle of operation is demonstrated. For the special case when the importance of delayed neutrons is negligibly small, the stability is demonstrated relative to arbitrary deviations from the equilibrium state.Translated from Atomnaya Énergiya, Vol. 21, No. 1, pp. 3–6, July, 1966.     

Dynamic model for control system design and simulation of a low temperature nuclear reactor

A lumped parameter dynamic model for the primary-loop and the U-tube steam generator of a low temperature power reactor is developed based on the fundamental conservation laws of fluid mass, energy and momentum. The dynamic model is formulated by coupling the point kinetics with reactivity feedback and the thermal-hydraulics of the reactor. The developed dynamic model is implemented on a personal computer using MATLAB/SIMULINK. Numerical simulation results for steady-state and transient responses are then presented, which show that the steady-state precision of the newly developed dynamic model is acceptable and the trend of the transient responses is correct. In addition, the “swell and shrink” behavior of the U-tube steam generator is also verified by numerical simulation. This newly established model can be utilized to control system design and simulation for the low temperature power reactor.