The importance of uncertainties in the ROP detector layout design process for CANDU reactors

In CANDU® reactor design, the regional overpower protection (ROP) systems protect the reactor against overpower in the fuel which could reduce the safety margin-to-dryout. The increase in fuel power could be caused by a localized power peaking within the core (for example, as a result of a certain reactivity device configuration) or a general increase in the core power level during a slow-loss-of-regulation (SLOR) event. This overpower could lead to fuel sheath dryout. In the CANDU® 600 MW (CANDU 6) design, there are two ROP systems in the core, one for each fast-acting shutdown system. Each ROP system includes a number of fast-responding, self-powered flux detectors suitably distributed throughout the core within vertical and horizontal assemblies. A new methodology for designing the detector layout for the ROP system, called the DETPLASA algorithm, has been developed recently. This method utilizes the simulated annealing (SA) technique to optimize the placement of the detectors in the core. The evaluation of the trip setpoint (TSP) corresponding to each detector layout configuration (i.e., each history in the SA algorithm) is performed probabilistically using the ROVER-F code. In this evaluation, there are uncertainties related to both the detector components (i.e., related to the margin-to-trip) and to the fuel channel components (i.e., related to the margin-to-dryout). In this paper, the importance of these uncertainties on the outcome of the detector layout optimization process is evaluated. Some parametric studies have been performed to quantify the effect of uncertainties on the resulting detector layout. Two types of investigations have been performed. First, a given detector layout will be used to explicitly determine the effect of changing the uncertainty values. In this study, 343 sets of uncertainty values are used to produce the corresponding TSP values. The variation in the TSP values is analyzed. Second, three sets of uncertainty values (a subset of uncertainties from the first study) are used in independent DETPLASA executions. The resulting detector layout configurations will be examined to observe the effect of these uncertainties on the final design. Results from these investigations are presented in this paper.     

Balancing human and technical reliability in the design of advanced nuclear reactors

The strong influence of human factors (HF) on the safety of nuclear facilities is nowadays recognised and the designers are now enforced to consider HF requirements in the design of new facilities. Yet, this consideration of human factors requirements is still more or less restricted to the latest phases of the projects, essentially for the design of human-system interfaces (HSI's) and control rooms, although the design options influencing at most the human performance in operation are indeed fixed during the very early phases of the new reactors projects.The main reason of this late consideration of HF is that there exist few methods and models for anticipating the influence of fundamental design options on the future performance of operation teams.This paper describes a set of new tools permitting (i) determination of the impact of the fundamental process design options on the future activity of the operation teams and (ii) assessment of the influence of these operational constraints on teams performance. These tools are intended to guide the design of future 4th generation (GEN4) reactors, within the frame of a global risk-informed design approach, considering technical and human reliability exigencies in a balanced way.     

Modeling and simulation of irradiation hardening in structural ferritic steels for advanced nuclear reactors

Hardening and embrittlement are controlled by interactions between dislocations and irradiation induced defect clusters. In this work we employ the visco plastic self consistent (VPSC) polycrystalline code in order to model the yield stress dependence in ferritic steels on the irradiation dose. We implement the dispersed barrier hardening model in the VPSC code by introducing a hardening law, function of the strain, to describe the threshold resolved shear stress required to activate dislocations. The size and number density of the defect clusters varies with the irradiation dose in the model. We find that VPSC calculations show excellent agreement with the experimental data set. Such modeling efforts can both reproduce experimental data and also guide future experiments of irradiation hardening.     

Numerical modeling and design of supercritical CO2 pre-cooler for fusion nuclear reactors

One of the key issues of fusion technology is the efficient recovery of the fusion power extracted by heat transfer fluids in the breeding blanket. The Spanish National Program TECNO_FUS is exploring a dual-coolant breeding blanket design concept and its plant auxiliary systems for a future power reactor (DEMO), with liquid lead–lithium as main primary nuclear power recovering fluid. Supercritical CO₂ is chosen for the secondary circuit, since its high efficiency at significantly lower required temperatures than for the Brayton helium cycle, due to low compression work near the critical point and also because its additional major benefits in terms of tritium control. Use of printed circuit heat exchangers (PCHE) is suggested in literature due to its highly compact design and robustness for the high pressures found. This work deals with the heat exchanger devoted to release the thermal energy of the power cycle to the thermal sink. The aim of this work is analyzing how the nearness of the CO₂ to its critical point affects the performance of the heat exchanger. Computer Fluid Dynamics (CFD) simulations that include the complex thermal behavior of CO₂ properties at supercritical conditions are used in order to achieve an accurate approach to the design of this heat exchanger. These results are compared with others obtained through correlations found in the open literature. The behavior of CO₂ close to its critical point results in an inefficient use of the exchange area, giving a temperature profile in CO₂ which remembers a condensation process and an overall heat transfer coefficient 1.4 times higher than the one achieved with literature correlations design.     

Fuel burn-up in nuclear reactors

The paper describes the method of calculating fuel burn-up in nuclear reactors, taking into account the capture and multiplication of neutrons while slowing down. In the calculations, account is taken of the burn-up of U²³⁵ and the build-up and burn-up of Np²³⁹, Pu²³⁹, Pu²⁴⁰, Pu²⁴¹ and of the fission fragments.     

Regulatory research in structural design and analysis of nuclear power plants

The paper gives a survey of research conducted or planned by the US Nuclear Regulatory Commission in the area of loading specification and structural mechanics as applied to safety analysis of nuclear power plant structures.     

Thermal shock in nuclear reactors

Most types of liquid cooled reactors undergo rapid changes in coolant temperature during abnormal operating conditions, which cause components adjacent to the coolant to undergo rapid surface temperature changes. The differential expansion between surface and bulk material caused by repeated thermal shocks can induce fatigue or cyclic creep damage. In thick walled components individual thermal shocks may cause rapid fracture from pre-existing defects. This paper reviews the reported experimental and analytical work defining the temperature, stress and resultant damage in reactor structural components subject to thermal shock.     

A fuzzy controller design for nuclear research reactors using the particle swarm optimization algorithm

In this paper, a closed-loop fuzzy logic controller based on the particle swarm optimization algorithm is proposed for controlling the power level of nuclear research reactors. The principle of the fuzzy logic controller is based on the rules constructed from numerical experiments made by means of a computer code for the core dynamics calculation and from human operator's experience and knowledge. In addition to these intuitive and experimental design efforts, consequent parts of the fuzzy rules are optimally (or near optimally) determined using the particle swarm optimization algorithm. The contribution of the proposed algorithm to a reactor control system is investigated in details. The performance of the controller is also tested with numerical simulations in numerous operating conditions from various initial power levels to desired power levels, as well as under disturbance. It is shown that the proposed control system performs satisfactorily under almost all operating conditions, even in the case of very small initial power levels.     

Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.     

Fuel elements of nuclear reactors

On the basis of foreign reports presented at the Second International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1958, the characteristics of construction of fuel elements (FE) and basic data relating to them for a series of reactors are given. Problems on the selection of fuel and construction materials as well as the technology of preparing FE for various types of nuclear reactor are examined.     

Trial decommissioning of nuclear reactors

Translated from Atomnaya Énergiya, Vol. 69, No. 3, pp. 157–160, September, 1990.     

Electrical simulation of nuclear reactors

The application of electrical simulation methods to nuclear reactor calculations considerably shortens the time required for the computational work. In this review the advantages and disadvantages of simulators are discussed and an example is given of the simulation of the reactor isotopic composition with the help of the simulator MN-7. The effectiveness of the electrical simulation method for the investigation of non-stationary reactor processes is shown by an example of design calculations carried out for an automatic power regulator for a reactor. It is shown that it is possible to simulate nonstationary processes occurring in a power reactor, taking into account the temperature coefficients of the reactivity; other applications of simulators in reactor design are indicated.     

Heating of the structures surrounding nuclear reactors

In the paper is considered the question of the effect of heating and radioactive radiation from a nuclear reactor on the concrete biological shield surrounding it. Various types of concrete used for the indicated purpose are described.Experimental data are given for the mechanical properties and for favorable mixes for the concrete of biological shields, which withstand heating and radiation in working reactors, and data are also given for materials used for heat shielding of nuclear reactors and methods of cooling the biological shield.     

Very high temperature design criteria for nuclear heat exchangers in advanced high temperature reactors

This review summarizes the results obtained in a research programme sponsored by the Ministry of Interior of the FRG on fundamental principles of a structural design code for heat exchanging components in nuclear process heat plants. Materials selection and design data, methods for dimensioning and the limitation of stress, strain, creep fatigue and buckling as well as environmental effects are discussed.