Analytical modeling of flashing-induced instabilities in a natural circulation cooled boiling water reactor

A dynamic model for natural circulation boiling water reactors (BWRs) under low-pressure conditions is developed. The motivation for this theoretical research is the concern about the stability of natural circulation BWRs during the low-pressure reactor start-up phase. There is experimental and theoretical evidence for the occurrence of void flashing in the unheated riser under these conditions. This flashing effect is included in the differential (homogeneous equilibrium) equations for two-phase flow. The differential equations were integrated over axial two-phase nodes, to derive a nodal time-domain model. The dynamic behavior of the interface between the one and two-phase regions is approximated with a linearized model. All model equations are presented in a dimensionless form. As an example the stability characteristics of the Dutch Dodewaard reactor at low pressure are determined.     

Core power limits for a lead-bismuth natural circulation actinide burner reactor

The Idaho National Engineering and Environmental Laboratory and Massachusetts Institute of Technology are investigating the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The design being considered here is a pool type reactor that burns actinides and utilizes natural circulation of the primary coolant, a conventional steam power conversion cycle, and a passive decay heat removal system. Thermal-hydraulic evaluations of the actinide burner reactor were performed to determine allowable core power ratings that maintain cladding temperatures below corrosion-established temperature limits during normal operation and following a loss-of-feedwater transient. An economic evaluation was performed to optimize various design parameters by minimizing capital cost. The transient power limit was initially much more restrictive than the steady-state limit. However, enhancements to the reactor vessel auxiliary cooling system for transient decay heat removal resulted in an increased power limit of 1040 MWt, which was close to the steady-state limit. An economic evaluation was performed to estimate the capital cost of the reactor and its sensitivity to the transient power limit. For the 1040 MWt power level, the capital cost estimate was 49 mills per kWhe based on 1999 dollars.     

Reliability assessment of emergency exhaust system in a pool-type research reactor

The reliability of an extract system in a swimming-pool-type research reactor has been assessed. A global fault-tree analysis technique has been utilized. The basic event reliability data is based on both generic and reactor specific informations.The unavailability of the extract system is quantified in terms of the unavailability of the various functional requirements of the system. The unavailability is expressed as the probability of failure on demand. The computer system unavailability is determined from the minimal cutsets of the system. It is found that only three events have a major contribution to the top event, i.e., failures of compressed air supply, electric power supply and solenoid valve. A sensitivity analysis is performed to show the effects of variations in the data values of the dominant cutsets. An uncertainty analysis was also performed on the fault tree. The evaluations show that the reactor extract system lacks diversity and redundancy in most of its components. It is tolerant of most minor degradations, as these are taken care of by the operating policies and procedures. However, it can not tolerate common cause failures, e.g., simultaneous compressed air and electric power supply failure. Based upon the results obtained some recommendations are made.     

Study on the flow-pattern-transition instability in a natural circulation heavy water moderated boiling light water cooled reactor

A mathematical model has been developed to study the flow pattern transition instability which may occur in a boiling two-phase system. The model considers flow pattern transition criteria for vertical upward and horizontal flow in pipes to identify the flow pattern transition and flow pattern specific pressure drop models. It also considers the drift flux model to estimate the void fraction in the two-phase region. The model has been applied to predict the flow pattern transition instability in a natural circulation heavy water moderated boiling light water cooled reactor. It is found that the instability characteristics is similar to that of the Ledinegg-type instability. However, the number of multiple steady states for a given operating power can be much larger in the flow pattern transition instability as compared to that of the Ledinegg-type instability. Stability maps were plotted and compared for both the flow pattern transition instability and that of the Ledinegg-type instability. The influence of various geometric and operating parameters on this instability were investigated.     

Gas cooled fast reactor research in Europe

Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce.GFR research within Europe is performed directly by those states who have signed the “System Arrangement” document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with the Generation IV GFR system.     

Reactivity insertion mechanisms in the gas-cooled fast breeder reactor

This report summarizes an analysis of reactivity insertion mechanisms in the gas-cooled fast breeder reactor (GCFR). Inherent reactivity feedback mechanisms are identified and their effects on reactor start-up, during normal operation, and on anticipated and postulated transients are analyzed. Potential sources of accidental reactivity insertions and the resulting transients are investigated, including potential reactivity effects due to cladding and fuel melting. All nuclear calculations are based on the ENDF-B, Version 3, cross-section file. It is concluded from these analyses that the GCFR is an inherently stable reactor during start-up and normal operation. Potential accidental reactivity insertions are mild, and in each case the reactor can be controlled with a substantial margin for fuel melting or cladding damage. In low-probability accident sequences which lead to core melting, there are potential fuel motion mechanisms which can mitigate reactivity effects and accident consequences.     

Natural circulation capability of Pb-Bi cooled fast reactor: PEACER

First-principle calculations were performed to analyze the natural circulation heat removal from the core of a liquid metal reactor (LMR). The lead-bismuth (Pb-Bi) was chosen as the primary coolant for the LMR system. From the single channel analysis the temperature and the pressure drop are calculated along the fuel assembly. The total pressure drop of the core is less than 100kPa due to the large pitch-to-diameter ratio and the small height of the fuel pin. The natural circulation potential is a key characteristics of the LMR design. The steady-state momentum and energy equations are solved along the primary coolant path. The calculations are divided into two parts: an analytical model and a one-dimensional lumped parameter flow loop model. Results of the analytical model indicate that the elevation difference of 4.5m between thermal centers of the core and the steam generators could remove as much as 10% of the nominal operating reactor power. The flow loop model yielded the total pressure drop and the natural circulation heat removal capacity.     

Effects of ship motions on natural circulation of deep sea research reactor DRX

DRX is a very small integral type PWR (750 kW_t) for a scientific deep-sea research bathyscaph. The core having a small amount of steam is cooled by natural circulation and pressurized by self-pressurization. During operation of the bathyscaph in a deep sea or near the water surface, a ship inclination or ship motions will affect the reactor behavior. This paper describes the effect of a heeling or a heaving on the thermal hydraulic behavior of reactor system, which is analyzed by improved so to simulate the effect of ship motions. The dynamics has a feature of nuclear power-natural circulation flow coupling under the condition of external forces. The analysis shows that ship inclination induces the core flow to decrease but reactor power recovers to the initial level without help of the reactor automatic control system. The heaving makes the core flow and the reactor power oscillate in phase with heaving, which are different from a density wave oscillation. Oscillation amplitudes of the flow and the power have peaks at the heaving period of 5 s. The peaks are due to resonance of the natural circulation flow and the heaving. An effective measure to suppress this oscillations due to heaving is to pressurize the primary loop by filling non-condensable gas. The density wave oscillation occurs when the reactor power increases over the rated power, and the boundary of its occurrence is analytically revealed. Under the condition of both density wave oscillation and heaving, the system shows to oscillate with the overlapped effect.     

High burn-up capability possibilities for a new beryllium moderated water cooled natural uranium reactor

A recently conceived nuclear reactor design is evaluated here for theoretical burn-up characteristics which might support Global Nuclear Energy Partnership (GNEP) goals. This reactor uses natural uranium metal as fuel with beryllium moderation. The reactor also uses light water as a coolant. The reactor analysis in this work predicts the reactor to be capable of running at up to 4 GW-thermal for total burn-up values of approximately 1.4 × 10³ GW-days. This is a very simple conceptual reactor design intended solely for very preliminary feasibility studies.     

Flow of kinetic parameters in a typical swimming pool type research reactor

Calculations were performed to estimate the variation in kinetic parameters (delayed neutron fraction and prompt neutron generation time) in different core configurations of a typical swimming pool type research reactor. Pakistan research Reactor-1 (PARR-1) was employed for this study. The effect due to burnup of the core was also studied. Calculations were performed with the help of computer codes WIMSD/4 and CITATION. Precursors yield was modified according to the neutron flux averaging only. This is the simple way to calculate the precursor yield for a particular core. The kinetic parameters are different for different core configurations. The β_eff decreases with 1.33 × 10⁻⁶/% burnup whereas prompt neutron generation time increases with 6.42 × 10⁻⁸ s/% burnup. The results were compared with safety analysis report and with published values and were found in good agreement. This study provides the confidence to understand the change in the kinetic parameters of research reactors with core change and also with burnup of the core.     

Improvement in understanding of natural circulation phenomena in water cooled nuclear power plants

The IAEA has organized a coordinated research project (CRP) on “Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems That Utilize Natural Circulation.” Specific objectives of CRP were to (i) establish the status of knowledge: reactor start-up and operation, passive system initiation and operation, flow stability, 3-D effects, and scaling laws, (ii) investigate phenomena influencing reliability of passive natural circulation systems, (iii) review experimental databases for the phenomena, (iv) examine the ability of computer codes to predict natural circulation and related phenomena, and (v) apply methodologies for examining the reliability of passive systems. Sixteen institutes from 13 IAEA Member States have participated in this CRP. Twenty reference advanced water cooled reactor designs including evolutionary and innovative designs were selected to examine the use of natural circulation and passive systems in their designs. Twelve phenomena influencing natural circulation were identified and characterized: (1) behaviour in large pools of liquid, (2) effect of non-condensable gases on condensation heat transfer, (3) condensation on the containment structures, (4) behaviour of containment emergency systems, (5) thermo-fluid dynamics and pressure drops in various geometrical configurations, (6) natural circulation in closed loop, (7) steam liquid interaction, (8) gravity driven cooling and accumulator behaviour, (9) liquid temperature stratification, (10) behaviour of emergency heat exchangers and isolation condensers, (11) stratification and mixing of boron, and (12) core make-up tank behaviour. This paper summarizes the achievements within the CRP for the first five phenomena (1-5).     

Reactivity effects in a BN-350 reactor

Conclusions The experiments were the first check on values of the temperature and power reactivity effects and the depletion effect for a fast power reactor. These results and their theoretical analysis enable us to estimate the reactivity balance in a BN-350 reactor, and to check and refine the methods of calculation. The models and methods used for calculations in designing fast power reactors turned out to be effective and gave quite satisfactory results.Calculations of the temperature and power reactivity effects somewhat underestimate their values (by 15%). The largest error in the temperature effect comes from the Doppler and sodium components, and in the power effect from the Doppler effect and the indeterminacy in temperature field calculations.Calculations give a fairly good estimate of the depletion effect also, although in this case obvious care has to be exercised in choosing the method of calculation. This fact is connected with the necessity of correctly allowing for nonuniformity of fuel depletion and the accumulation of plutonium.Investigation of the reactivity effect for prolonged use of the reactor is of great practical and theoretical interest. The effect of depletion and change in structure of the fuel pellets on the temperature and power effects, the nonlinear nature of the power effect, and the influence of plutonium accumulation on the reactivity effects in the reactor, all these are questions of reactor physics the study of which allows us to increase the accuracy of physical calculations in designing fast power reactors.Translated from Atomnaya Énergiya, Vol. 42, No. 1, pp. 3–8, January, 1977.     

Subchannel analysis of a fast reactor cooled by supercritical light water

The concept of a high temperature fast reactor cooled by supercritical water (SCFR-H) was developed for achieving high thermal efficiency and a compact reactor system. The core characteristics were obtained from single channel thermal-hydraulic analysis. Thus, it is necessary to carry out subchannel analysis to estimate the effect of local power peaking and cross flows. For this purpose, a subchannel analysis code is developed. It is verified by comparing the results with experimental data of High Conversion Pressurized Water Reactor (HCPWR). Sensitivities of the outlet coolant and cladding temperature to the subchannel flow area and local power peaking are high. One of the reasons is that the ratio of the coolant flow rate of SCFR-H to the power is smaller than that of LWR. Another reason is that, temperature of supercritical water is more sensitive to the enthalpy change above 450°C. The outlet coolant temperature distribution can be flattened by reducing the area of the peripheral subchannels and by enhancing the mixing between the subchannels.     

Optimization of a radially cooled pebble bed reactor

By altering the coolant flow direction in a pebble bed reactor from axial to radial, the pressure drop can be reduced tremendously. In this case the coolant flows from the outer reflector through the pebble bed and finally to flow paths in the inner reflector. As a consequence, the fuel temperatures are elevated due to the reduced heat transfer of the coolant. However, the power profile and pebble size in a radially cooled pebble bed reactor can be optimized to achieve lower fuel temperatures than current axially cooled designs, while the low pressure drop can be maintained.The radial power profile in the core can be altered by adopting multi-pass fuel management using several radial fuel zones in the core. The optimal power profile yielding a flat temperature profile is derived analytically and is approximated by radial fuel zoning. In this case, the pebbles pass through the outer region of the core first and each consecutive pass is located in a fuel zone closer to the inner reflector. Thereby, the resulting radial distribution of the fissile material in the core is influenced and the temperature profile is close to optimal.The fuel temperature in the pebbles can be further reduced by reducing the standard pebble diameter from 6 cm to a value as low as 1 cm. An analytical investigation is used to demonstrate the effects on the fuel temperature and pressure drop for both radial and axial cooling.Finally, two-dimensional numerical calculations were performed, using codes for neutronics, thermal-hydraulics and fuel depletion analysis, in order to validate the results for the optimized design that were obtained from the analytical investigations. It was found that for a radially cooled design with an optimized power profile and reduced pebble diameter (below 3.5 cm) both a reduction in the pressure drop ( bar), which increases the reactor efficiency with several percent, and a reduction in the maximum fuel temperature (C) can be achieved compared to present axially cooled designs.     

Modeling aspects in linear stability analysis of a self-pressurized, natural circulation integral reactor

The stability of a self-pressurized natural circulation integral reactor is studied by means of a linear approach, taking the CAREM-25 reactor as reference.A thermohydraulic code has been improved for analysis of linear stability, great emphasis having been placed on the minimization of numerical diffusion and integration errors. A linearization method is implemented by means of numerical perturbations. The results are obtained within the frequency domain. The code is compared to a simpler analytical model, by contrasting stability maps obtained from both models for a test configuration, showing good agreement.In this type of reactor, oscillations are promoted by the two-phase regime in its long riser, and take place due to the counteraction between mass flow and buoyancy force.The stability of the system is strongly influenced by the steam-dome dynamics. Condensation in the steam zone, together with reactor power, determines the dynamical state of the system.The phase-lag introduced by the core dynamic regarding the riser timing, together with the sensitivity of the buoyancy force due to flow changes, determines the sustainability of the oscillation. A parametric study is carried out, gradually increasing the complexity of the model, to analyze the influence of different factors on the oscillation sustainability, concerning physical process and modeling approaches. The analysis includes the relative velocities between phases, the axial power profile along the core, the buoyancy force due to subcooled density changes, the flashing effect, the core dynamic and the pressure feedback due to self-pressurization. The steam-dome-pressure feedback is identified as a stabilizing effect, as long as it decreases the sensitivity of the buoyancy force.     

Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor

The main aim of this work is to identify how much the code results are affected by the code user in the choice of, for example, the number of thermal hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previously validated nodalization for analysis of steady-state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of modeling. The results highlight the necessity of sensitivity analysis to obtain the ideal modeling to simulate a specific system.     

Potential and limits of water cooled divertor concepts based on monoblock design as possible candidates for a DEMO reactor

In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term (“conservative baseline design”). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies.For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed.Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R&D activities are reported.This work has been carried out in the frame of EFDA PPPT Work Programme activities.