Stability analysis of a liquid-metal reactor and its primary heat transport system

Part of the reactor design process is the assessment of the impact of different design changes on pre-defined performance criteria including stability of the reactor system under different conditions. This work focuses on the stability analysis of a combined liquid-metal reactor and primary heat transport system where system parameters are free to vary, with particular interest in low reactor power, low reactor coolant flow conditions. Such conditions might be encountered, for example, after a loss of flow without scram in some passively safe reactor designs. Linear-stability-analysis-based methods are developed to find the stability regions, stability boundary surface in system parameter space, and frequency of oscillation at oscillatory instability boundaries. Models are developed for the reactor, detailed thermal hydraulic reactivity feedback associated with coolant outlet and inlet temperatures, decay heat and primary system. The developed stability analysis tools are applied to the system model. The system parameters include integral reactivity parameters, decay heat, primary system mass, coolant flow and natural circulation flow. The resulting stability boundary surface and its associated frequency of oscillation surface in multidimensional system parameter space show the effect of system parameter changes. By adopting model parameters from liquid-metal reactor designs, a stability prediction procedure is illustrated.     

TACR热传输系统热工水力瞬态分析

钍燃料的利用对于缓解核燃料资源短缺具有重要意义,坎杜型反应堆(Canadian Deuterium Uranium,CANDU)在堆芯布置、中子利用效率及先进燃料循环方面具有较高的灵活性,使得其在CANDU反应堆中引入钍燃料循环更具现实意义。CANDU型反应堆中钍基燃料应用关键基础技术研究是加拿大与我国正在开展的合作课题,其中开发自主的CANDU堆堆芯热工水力设计和安全分析程序是钍基燃料应用必不可少的设计工作之一。本文针对CANDU型反应堆热传输系统结构特点,采用FORTRAN程序设计语言开发了适用于CANDU型反应堆热传输系统的热工水力瞬态分析程序CANTHAC(CANDU Thermal-Hydraulic Analysis Code)。利用CANTHAC对钍基先进CANDU堆(Thorium-based Advanced CANDU Reactor,TACR)进行了瞬态分析,计算工况包括满功率稳态、无保护蒸汽发生器(Steam Generator,SG)二次侧给水温度降低事故及完全失流事故。其中,满功率稳态计算结果与清华大学设计的钍基先进CANDU堆TACR设计值吻合较好,相对误差不超过2%,在可接受范围内;无保护SG二次侧给水温度降低事故及完全失流事故在计算条件下所得的燃料温度及系统压力等关键热工水力参数均在安全限值内,满足安全准则要求。程序为模块化编程,便于移植和改进,具有一定的通用性,为进一步研究工作奠定了基础。     

Comparative forced vibration test of two BWR-type reactor buildings

The basis of this paper is comparative forced vibration testing of two GE 460 MW(e) BWR-type reactor buildings. The tested nuclear power plants are the Fukushima Nuclear Power Plant Unit No. 1 of the Tokyo Electric Power Company (hereinafter referred to as Fukushima) and the Shimane Nuclear Power Plant of the Chugoku Electric Power Company (Shimane). They are almost the same in both structure and function, but are built on rock of quite different rigidity. The shear wave velocity of Shimane is about three times that of Fukushima. The forced vibration tests were performed immediately after completion of each reactor building using a vibrator with a maximum exciting force of 3 t. The computer simulation analyses were carried out using vibration models possessing different internal viscous damping factors for each structural element. Both the resonance periods and damping factors of Fukushima were found to be larger than those of Shimane. Thus, site selection of nuclear power plants must be reviewed as a matter of utmost importance from the viewpoint of seismic design.     

Kinetic parameters of a low enriched uranium fuelled material test research reactor at end-of-life

The kinetic parameters at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA’s 10 MW benchmark reactor. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time and effective delayed neutron fraction. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that in comparison with the beginning-of-life values, at end-of-life, the neutron flux increased throughout the core, the prompt neutron generation time increased by 3.68% while the effective delayed neutron fraction decreased by 0.35%.     

Neutron transport in a reactor system with temperature feedback

When the effect of temperature feedback in a reactor system is considered the neutron transport equation for the neutron density is supplemented by a temperature equation which is a partial differential equation of parabolic type if heat conduction is taken into consideration. This consideration leads to a coupled system of nonlinear partial integro-differential equations. The aim of this paper is to present an iterative scheme for the determination of the solution of the nonlinear coupled system and to establish some qualitative property of the solution. The iterative scheme consists of two monotone sequences which converge monotonically from above and below, respectively, to a unique solution. The qualitative aspect includes the existence and uniqueness of a positive solution, upper and lower bounds of the solution and stability of a steady-state solution.     

Reactivity feedback coefficients of a low enriched uranium fuelled material test research reactor at end-of-life

The reactivity feedback coefficients at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA’s 10 MW benchmark reactor. Simulations were carried out to calculate the different reactivity feedback coefficients including Doppler feedback coefficient, reactivity coefficient for change of water temperature and reactivity coefficient for change of water density. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitude of all the reactivity feedback coefficients increased at end of life of the reactor by almost 2–5%.     

Utilization of natural uranium in a homogeneous reactor

One of the main advantages of homogeneous reactors over heterogeneous reactors is the absence of stability problems with heat-emitting elements. Therefore, the permissible degree of uranium utilization, and consequently, the economic characteristics of a homogeneous reactor can be significantly improved.This paper indicates the magnitude of uranium utilization with respect to the initial system parameters and establishes conditions for the probability of capture by U²³⁸ of slowed neutrons, the occurrence of which permits maximum utilization in a reactor using regenerated uranium.     

SIMMER model of a low-enriched uranium non-power reactor

IRSN has started using the coupled neutronics–fluid dynamics code SIMMER [Tobita, Y., Kondo, Sa., Yamano, H., Morita, K., Maschek,W., Coste, P., Cadiou, T., 2006. The development of SIMMER-III, an advanced computer program for LMFR safety analysis, and its application to sodium experiments. Nucl. Technol. 153 (3), 245] to study core-disruptive accidents induced by insertions of large reactivities to produce very short period power excursions in fuel plate-type and water-moderated experimental research reactors. Until now, French safety analyses retain a bounding thermal energy released and mechanical yields, deduced from analysis of destructive in-pile test programs, to study the behavior of such reactors and design their structures and containment.Contrary to this approach, the present research program aims at modeling the design basis accident of research reactors with a low-enriched fuel using a CFD code. The objective is to analyze the effects of reactivity feedbacks and how they would limit the generated thermal energy released in the fuel. These aspects require a close coupling of the neutronics to the fluid dynamics analysis. The consequences of the nuclear power excursion, the changes of state of the fuel and the coolant, and ultimately the mechanical energy released are calculated by SIMMER. For large step-wise reactivity introductions, the Doppler effect and, at a lower extent, the fuel element thermal dilatation, which generates locally a decrease of the moderator to fuel ratio, limit the power excursion before the energy released is high enough to melt a large part of the fuel. Moreover, it has been shown that imposing an external reactivity as a step-wise or time-dependent reactivity introduction yields results quite different from those of the physical movement of control rods.     

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl_x-Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si-Al and U₃Si₂-Al) and oxide (U₃O₈-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U₃Si₂-Al followed by 0.03% for U₃Si-Al, and 0.01% for U₃O₈-Al fuel. The U₃O₈-Al fueled reactor gave the maximum ρ_excess at BOL which was 21.67% more than the original fuel followed by U₃Si-Al which was 2.55% more, while that of U₃Si₂-Al was 2.50% more than the original UAl_x-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U₃O₈-Al followed by U₃Si-Al and then U₃Si₂-Al fuel.     

Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl_x–Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si–Al and U₃Si₂–Al) and oxide (U₃O₈–Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 °C to 50 °C and 100 °C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U₃O₈–Al was about 2% more than the original UAl_x–Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.     

Acoustic characterization of a CANDU primary heat transport pump at the blade-passing frequency

In this paper, we examine the acoustics of a single-stage, double-volute CANDU heat transport pump based on a full-scale experimental investigation. We estimate the strength of source variables (acoustic pressure and velocity) and establish the pump characteristics as an acoustic source at the blade-passing frequency. We conduct this analysis by first assessing the resonance effects in the test loop, and then decomposing the measured signal into the components associated with pump action and loop acoustics with the use of a simple pump model. The pump model is based on a linear superposition of pressure wave transmission and excitation. The results of this analysis indicate that the pump source variables are nearly free of acoustic resonance effects in the test loop. The source pressure and velocity are each estimated at approximately 10 kPa (zero-to-peak). The results also indicate that the pump may act as both a pressure and a velocity source. At the loop resonance, the pump acoustic behavior is exclusively governed by the pressure term. This observation leads to the conclusion that the maximum amplification of pressure pulsations in a reactor heat transport system may be predicted by modeling the pump as a pressure source.     

A review of radionuclide behaviour in the primary system of a very-high-temperature reactor

During normal operation of (V)HTRs radiologically-significant contamination of the primary system will occur this being of prime importance in relation to depressurization accidents. This paper reviews information relevant to radiocontaminant transport in (V)HTR primary systems paying particular attention to chemical forms, interactions with dust and overall distribution. The primary-system environment comprises nuclear graphites, alloys, dust and high-purity helium into which low releases of the isotopes ¹³⁴Cs, ¹³⁷Cs, ⁹⁰Sr, ^110mAg, ¹³¹I, ¹³⁵Xe and ⁸⁵Kr can be anticipated. Additionally, proper treatment of radiological risk requires accounting for tritium.A likely gas-phase speciation of the chemically-active fission products is proposed:

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for caesium and strontium, hydroxides would be dominant with iodides as minor species if a relatively low iodine concentration can be assumed;

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for iodine, a split between CsI and HI are likely to dominate with atomic iodine as a minor species.

Strong sorption of radionuclides onto carbonaceous dust can be expected. Such dust is likely to cover all surfaces in a pebble-bed (V)HTR so radionuclides will principally associate with this dust rather than underlying alloys. This is unlikely in prismatic (V)HTRs with lower and uneven dust deposits. Where caesium interacts with alloys strong implanting of a large fraction can occur via adsorption and reaction with low-concentration silicon. Silver shows no special affinity for carbonaceous dust but may interact preferentially with nickel-rich alloys, i.e., in the IHX and/or the gas turbine. Quantitative evaluations of radionuclide distribution are hampered by a lack of data regarding sorption onto the graphites, alloys and carbonaceous dust of modern (V)HTR systems; a long time will elapse before sufficient data are forthcoming. In the meantime, some form of best-estimate distribution and upper-bound concentration for contamination is needed if deterministic safety evaluations are to begin. This distribution will be different for pebble-bed and prismatic designs.     

Effects of high density dispersion fuel loading on the dynamics of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the dynamics of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different properties affecting the reactor in different ways, fuels U–Mo (9w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to determine the reactor performance under reactivity insertion and loss of flow transients. Nuclear reactor analysis code PARET was employed to carry out these calculations. It is observed that during the fast reactivity insertion transient, the maximum reactor power is achieved and the energy released till the power reaches its maximum increases by 45% and 18.5%, respectively, as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved during the transient, by 27.7 K, 19.7 K and 7.9 K, respectively. The time required to reach the peak power decreases. During the slow reactivity insertion transient, the maximum reactor power achieved increases slightly by 0.3% as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³ but the energy generated till the power reaches its maximum decreases by 5.7%. The temperatures of fuel, clad and coolant outlet remain almost the same for all types of fuels. During the loss of flow transients, no appreciable difference in the power and temperature profiles was observed and the graph plots overlapped each other.     

Modeling of turbulent condensation heat transfer in the boiling water reactor primary containment

A condensation heat transfer model is developed for the purpose of predicting the atmosphere temperature response within the primary containment of a boiling water reactor during the initial forced convection heat transfer period following a postulated loss-of-coolant accident. The model utilizes simultaneous heat and mass transfer for the process of condensation in the presence of a non-condensible gas. The gas-vapor diffusion layer formed is in the mode of turbulent, forced convection. The predicted heat transfer is determined to be diffusion controlled with negligible resistance being contributed by the condensate film. The model is qualified through the analysis of the response of a containment test facility; the results compare favorably with experimental observations made by the General Electric Co. Predicted temperature responses for a typical containment are also shown and compared with those obtained through use of the Uchida heat transfer correlations.     

Development of a new decay heat removal system for a high temperature gas-cooled reactor

The heat removal capacity of a RCCS is one of the major parameters limiting the capacity of a HTGR based on a passive safety system. To improve the plant economy of a HTGR, the decay heat removal capacity needs to be improved. For this, a new analysis system of an algebraic method for the performance of various RCCS designs was set up and the heat transfer characteristics and performance of the designs were analyzed. Based on the analysis results, a new passive decay heat removal system with a substantially improved performance, LFDRS was developed. With the new system, one can have an expectation that the heat removal capacity of a HTGR could be doubled.     

A study of the sensitivity of loca heat transfer analysis for a water-cooled reactor system

Results are given of computer calculations, using the reactor thermal analysis code THETA1-B, to determine the significance and relative importance of various heat transfer regimes in predicting maximum fuel cladding temperature for the blowdown phase of a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor system. The factors considered include the choice of heat transfer correlation for a particular heat transfer regime, the method of delineating the boundaries between regimes, and core inlet coolant flow conditions.For a hot-leg rupture, the maximum surface temperature is sensitive to a number of factors, including choices of critical heat flux correlation, flow boiling transition heat transfer correlation, and in particular, stable film flow boiling correlation. However, for a LOCA resulting from a double-ended rupture of an inlet feeder, these factors have only marginal effects on maximum cladding temperature. In this case the importance of heat transfer to dry steam coolant at low net flow rate conditions is demonstrated, indicating a need for further information.     

Microstructure and elemental distribution of americium-containing uranium plutonium mixed oxide fuel under a short-term irradiation test in a fast reactor

In order to investigate the effect of americium addition in MOX fuel on the irradiation behavior, the ‘Am-1’ program is being conducted in the experimental fast reactor Joyo. The Am-1 program consists of two short-term irradiation tests of 10 min and 24 h irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. This paper reports on the results of PIEs for Am-containing MOX fuel irradiated for 10 min. MOX fuel pellets containing 3% or 5% Am were fabricated in a shielded air-tight hot cell using a remote handling technique. The oxygen to metal ratio (O/M) of these fuel pellets was 1.98. They were irradiated at peak linear heating rate of about 43 kW m⁻¹. Focus was being placed on migration behavior of Am during the irradiation. The ceramography results showed that structural changes such as lenticular pores and a central void occurred early, within the brief 10 min of irradiation. The results of electron probe microanalysis revealed that the concentration of Am increased in the vicinity of the central void.     

Design and analysis of a uranium hexafluoride actinide transmutation reactor

The long-term (> 1000 years) hazards of high-level wastes (HLW) can be reduced substantially by practising waste-actinide partitioning-transmutation (P-T). This paper investigates the waste-actinide transmutation performance of a uranium hexafluoride actinide transmutation reactor (UHATR). Using mostly present-day and near-term technology, a preliminary UHATR design is established. Because of the gaseous nature of the fuel, very high neutron fluxes are obtained. Compared with an LWR, the average blanket thermal flux of this UHATR is about 10–30 times higher, leading to a 15-fold improvement in the percentage of actinides fissioned per year of irradiation.     

Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U–Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.