Liquid Metal Fast Reactor Research Articles

Simulation of central sloshing experiments with smoothed particle hydrodynamics (SPH) method

Liquid sloshing phenomena can be observed whenever a liquid in a container has an unrestrained surface and can be excited. A particular type of sloshing motion can occur during the core meltdown of a liquid metal cooled reactor (LMR) and can lead to a compaction of the fuel in the center of the core possibly resulting in energetic nuclear power excursions. This phenomenon was studied in series of “centralized sloshing” experiments with a central water column collapsing inside the surrounding cylindrical tank. These experiments provide data for a benchmark exercise for accident analysis codes. To simulate “centralized sloshing” phenomena, a numerical method should be capable to predict the motion of the free surface of a liquid, wave propagation and reflection from the walls. In this study, a meshless method based on smoothed particle hydrodynamics (SPH) for the simulation of a 3D free surface liquid motion has been developed. The proposed method is applied to the simulation of “centralized sloshing” experiments. Simulation results are compared with the experimental results as well as with results of computations performed with the 3D code SIMMER-IV which is an advanced reactor safety analysis code that implements the traditional mesh-based numerical method. In a series of numerical calculations it is shown that overall motion of the liquid is in a good agreement with experimental observations. Dependence on the initial and geometrical symmetry is studied and compared with experimental data.

Effects of fuel type on the safety characteristics of a sodium cooled fast reactor. Part II: Simulation results

The objective of this investigation is to examine the impact of the fuel type on the inherent safety characteristics of Liquid Metal Fast Reactors (LMFRs). To perform this study, the responses to various transient conditions are examined for metallic, oxide and nitride cores of a baseline LMFR. GE-Hitachi’s Super Power Reactor Innovative Small Module (S-PRISM) was chosen as the baseline LMFR. In Part I of this paper, the background on S-PRISM’s metal and oxide cores are described and the redesign of a new nitride fueled S-PRISM core were introduced. Reactivity feedback and power profile data necessary for transient simulations with RELAP5-3D/ATHENA ( RELAP5-3D, 2009) code are also presented and discussed. In this Part II of our paper, we present the results of accident simulations and a comparison between the metal, oxide and nitride cores based on their performance during the selected accident scenarios. Loss of Flow, Loss of Heat Sink, Loss of Power and inadvertent control rod withdrawal accidents were simulated for each core at beginning (BOC), middle (MOC) and end of a fuel cycle (EOC). The simulations were stopped at the initiation of melting of fuel or cladding. The results showed that in most of the transients the metal core came closer to its melting temperature while the strong reactivity feedbacks of the oxide and nitride cores limited their fuel temperature increases. Overall, the oxide and nitride cores had similar performance with respect to their inherent safety characteristics.

Characterization of sodium flow over hexagonal fuel subassemblies

Steady flow of liquid sodium over a bundle of heat generating hexagonal subassemblies has been investigated. The cross flow pressure drop and heat transfer are characterized using the general purpose CFD code STAR-CD. Analysis has been carried out for both laminar and turbulent regimes of interest to liquid metal fast reactors. Turbulence has been modeled using low Reynolds number (Re) k– ε model. The estimated pressure drop and heat transfer coefficients are compared against that of a straight parallel plate channel. It is seen that in the low Reynolds number range, the pressure drop for the hexagonal path is nearly equal to that of the parallel plate channel for the same length. However, in the high Reynolds number range, the pressure drop of the hexagonal path is much higher than that in the parallel plate channel, the ratio being 2 at Re = 2000 while it is 3.6 at Re = 20,000. Two competing factors, viz., (i) jet impingement/flow development effect and (ii) flow separation effect are found to influence the average Nusselt number (Nu). In the laminar regime, the latter effect dominates leading to a decrease of the Nusselt number with an increase in the Reynolds number. However, in the turbulent regime, the former effect dominates leading to an increase in the Nusselt number with Reynolds number. The Nusselt number in the hexagonal path is about twice that of the parallel plate channel due to under development of velocity/temperature profiles and the recirculation associated with the hexagonal path due to the changes in flow direction. Detailed correlations for both the pressure drop and the average Nusselt number have been proposed.

Effects of fuel type on the safety characteristics of a sodium-cooled fast reactor: Part I: Background, modeling tools and pre-transient calculations

The objective of this investigation is to examine the impact of the fuel type on the inherent safety characteristics of Liquid Metal Fast Reactors (LMFRs). To perform this study, the responses to various transient conditions are examined for metallic, oxide and nitride cores of a baseline LMFR. GE-Hitachi’s Super Power Reactor Innovative Small Module (S-PRISM) was chosen as the baseline LMFR. In this paper the background on S-PRISM’s metal and oxide cores are described and the redesign of a new nitride fueled S-PRISM core is introduced. Reactivity feedback and power profile data necessary for transient simulations with RELAP5-3D/ATHENA code are also presented and discussed. Part II of this paper will present the results of accident simulations and a comparison between the metal, oxide and nitride cores based on their performance during the selected accident scenarios.

Reduction in Degree of Absorber-Cladding Mechanical Interaction by Shroud Tube in Control Rods for the Fast Reactor

Research and development of a long-life control rod for fast reactors is being conducted at Joyo. One of the challenges in developing a long-life control rod is the restraint of absorber-cladding mechanical interaction (ACMI). First, a helium-bonding rod was selected as a control rod for the experimental fast reactor Joyo, which is the first liquid metal fast reactor in Japan. Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B4C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B4C pellets and the cladding tube. However, once B4C pellets cracked and broke into small fragments, relocation occurred. After this, the narrow gap closed immediately as the degree of B4C pellet swelling increased. To solve this problem, the gap was widened during design, and sodium was selected as the bonding material instead of helium to restrain the increase in pellet temperature. Irradiation testing of the modified sodium-bonding control rod confirmed that ACMI would be restrained by the shroud tube regardless of the occurrence of B4C pellet relocation. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of postirradiation examination are reported.

INHERENT SAFETY ANALYSIS OF THE KALIMER UNDER A LOFA WITH A REDUCED PRIMARY PUMP HALVING TIME

The 600 MWe, pool-type, sodium-cooled, metallic fuel loaded KALIMER-600 (Korea Advanced LiquId MEtal Reactor, 600 MWe) has been conceptually designed with an emphasis on safety by self-regulating (inherent/intrinsic) negative reactivity feedback in the core. Its inherent safety under the ATWS (Anticipated Transient Without Scram) events was demonstrated in an earlier study. Initiating events of an HCDA (Hypothetical Core Disruptive Accident), however, also need to be analyzed for assessment of the margins in the current design. In this study, a hypothetical triple-fault accident, ULOF (Unprotected Loss Of Flow) with a reduced pump halving time, is investigated as an initiator of a core disruptive accident. A ULOF with insufficient primary pump inertia may cause core sodium boiling due to a power-to-flow mismatch. If the positive sodium reactivity resulting from this boiling is not compensated for by other intrinsic negative reactivity feedbacks, the resulting core power burst would challenge the fuel integrity. The present study focuses on determination of the limit of the pump inertia for assuring inherent reactivity feedback and behavior of the core after sodium boiling as well. Transient analyses are performed with the safety analysis code SSC-K, which now incorporates a new sodium boiling model. The results show that a halving time of more than 6.0 s does not allow sodium boiling even with very conservative assumptions. Boiling takes place for a halving time of 1.8 s, and its behavior can be predicted reasonably by the SSC-K.

Feasibility study of the application of exotic pin for tight-lattice fuel assembly

A tight-lattice fuel assembly having less space for the coolant is more feasibly applied in Liquid Metal Fast Breeder Reactor (LMFBR). The thermal hydraulic constraint due to smaller coolant space can be compensated by the high heat capacity of the liquid metal coolant. A tight pin configuration provides high fuel volume fraction which eventually gives better neutronic performance for longer core lifetime. A cylindrical pin array provides less flexible arrangement for tight-lattice assembly, which results in very narrow coolant gaps connecting its neighboring subchannels. Therefore, the so-called exotic pin shape is introduced, which enable to distribute the coolant flow more uniformly, to be applied in tight-lattice bundles with sodium coolant. As Nusselt number and wall friction correlation are absent for this type of geometry, CFD calculations are performed by employing k–ε turbulent model.

高速炉の1次系流量減少時スクラム失敗事象（ULOF)遷移過程における上限エナジェティクスに関する検討

高速炉の1次系流量減少時スクラム失敗事象（ULOF)遷移過程における上限エナジェティクスに関する検討

「もんじゅ」定格運転での1次冷却系の放射線被ばく線量の予測解析

Radioactive corrosion products (CP) are main source of personal radiation exposure during maintenance without fuel-failure accident in the Liquid-Metal Fast Breeder Reactor (LMFBR) plants. In order to establish the techniques of radiation dose estimation for personnel, program system “DORE” has been developed. The DORE system is constructed by PSYCHE code and QAD code system. The density of each deposited CP of primary coolant system in MONJU was estimated by using the PSYCHE. Moreover, the QAD-CGGP2R code is applied to dose rate calculations for the primary coolant system in MONJU. The dose rate around primary piping system was visualized using AVS software. The predicted values were estimated to be saturated at 2-3 mSv/h in twenty years after the start of operation, and the dose rate reaches 4 mSv/h in domains near the IHX and the cold-leg piping. It has been assumed that the main radiation source is 54Mn in the IHX, primary pump and cold-leg piping region. On the other hand, it was indicated that the contribution to dose rate of the 60Co accounted for approximately 23% in the hot-leg piping region.

Reduction in Degree of Absorber-Cladding Mechanical Interaction by Shroud Tube in Control Rods for the Fast Reactor

Research and development of a long-life control rod for fast reactors is being conducted at Joyo. One of the challenges in developing a long-life control rod is the restraint of absorber-cladding mechanical interaction (ACMI). First, a helium-bonding rod was selected as a control rod for the experimental fast reactor Joyo, which is the first liquid metal fast reactor in Japan. Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B4C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B4C pellets and the cladding tube. However, once B4C pellets cracked and broke into small fragments, relocation occurred. After this, the narrow gap closed immediately as the degree of B4C pellet swelling increased. To solve this problem, the gap was widened during design, and sodium was selected as the bonding material instead of helium to restrain the increase in pellet temperature. Irradiation testing of the modified sodium-bonding control rod confirmed that ACMI would be restrained by the shroud tube regardless of the occurrence of B4C pellet relocation. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of postirradiation examination are reported.

3D Simulation of Solid-Melt Mixture Flow with Melt Solidification Using a Finite Volume Particle Method

Relocation and freezing of molten core materials mixed with solid phases are among the important thermal-hydraulic phenomena in core disruptive accidents of a liquid-metal-cooled reactor (LMR). To simulate such behavior of molten metal mixed with solid particles flowing onto cold structures, a computational framework was investigated using two moving particle methods, namely, the finite volume particle (FVP) method and the distinct element method (DEM). In FVP, the fluid movement and phase changes are modeled through neighboring fluid particle interactions. For mixed-flow calculations, FVP was coupled with DEM to represent interactions between solid particles and between solid particles and the wall. A 3D computer code developed for solid-liquid mixture flows was validated by a series of pure-and mixed-melt freezing experiments using a low-melting-point alloy. A comparison between the results of experiments and simulations demonstrates that the present computational framework based on FVP and DEM is applicable to numerical simulations of solid-liquid mixture flows with freezing process under solid particle influences.

Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on 239Pu, 235U, 238U

We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400–600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4–1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for 99Mo, 95Zr, 137Cs, 140Ba, 141,143Ce, and 147Nd. Modest incident-energy dependence exists for the 147Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by ∼5–7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except for 99Mo where the present results are about 4%-relative higher for neutrons incident on 239Pu and 235U. Additionally, our results illustrate the importance of representing the incident energy dependence of fission product yields over the fast neutron energy range for high-accuracy work, for example the 147Nd from neutron reactions on plutonium. An upgrade to the ENDF library, for ENDF/B-VII.1, based on these and other data, is described in a companion paper to this work.

Thermal-hydraulics analyses of ELSY lead fast reactor with open square core option

Among the next generation of nuclear reactors, well-known as Gen-IV, the LFR (Lead Fast Reactor) is one of the most promising advanced reactor able to comply the principles of sustainability, economics, safety and proliferation resistance. The ELSY project (European Lead-cooled SYstem), funded by the 6th European Framework Programme, aims at investigating the technical/economical feasibility of a high power lead fast reactor with waste transmutation capability. Several innovative design solutions have been proposed at system level and some of them regard the core region with open-assemblies that represents the reference option. To support the design phase and the safety assessment of ELSY, the thermal-hydraulic code RELAP5, modified to treat heavy liquid metals, has been taken into account. The paper deals with the development of the ELSY thermal-hydraulic and point kinetic model for RELAP5 focusing the attention at core assembly level to verify that the temperatures at nominal power conditions stay within the safety limits both in Beginning-of-Cycle (BOC) and in End-of-Cycle (EOC) conditions. Moreover, in order to have a first evaluation of the system behavior in accidental conditions, an Unprotected Loss-of-Flow Accident (ULOF) simulation at BOC has been analysed and discussed. List of acronyms ADS Accelerator Driven System BOC Beginning-of-Cycle DEC Design Extension Condition EFIT European Facility for Industrial Transmutation ELSY European Lead-cooled SYstem EOC End-of-Cycle FA Fuel Assembly FP6 6th Framework Programme Gen-IV Generation IV (four) GESA Gepulste ElektronenStrahlanlage (Pulsed Electron Beam Facility) HX Heat eXchanger IC Isolation Condenser LFR Lead Fast Reactor LMFR Liquid Metal Fast Reactor PDS-XADS Preliminary Design Studies on eXperimental ADS RVACS Reactor Vessel Air Cooling System ULOF Unprotected Loss-of-Flow W-DHR Water Decay Heat Removal

Low cycle fatigue properties of modified 9Cr-1Mo ferritic martensitic steel weld joints in sodium environment

Mod.9Cr-1Mo ferritic-martensitic steel is the material chosen for the steam generator of the Prototype Fast Breeder Reactor being built at Kalpakkam, India. The use of sodium as a heat transfer medium for Liquid Metal Fast Breeder Reactors (LMFBRs) necessitates a comprehensive understanding of the effects of dynamic sodium on the Low Cycle Fatigue (LCF) behaviour of structural components. Moreover welds being the weak links in any structure, it is necessary to evaluate the LCF behaviour of joints in sodium environment, more so in Mod.9Cr-1Mo steel because of the well established Type — IV cracking in this material. With this aim in view, a programme has been initiated to evaluate the LCF properties of weld joints of this steel in dynamic sodium environment. A facility has been developed in-house for mechanical property evaluation in dynamic sodium. LCF tests conducted in flowing sodium environment at 823 and 873 K showed a similar trend in cyclic stress response in air and sodium environments exhibiting a continuous cyclic softening behaviour. The fatigue lives are significantly improved in sodium environment when compared to the data obtained under identical testing conditions in air environment. The lack of oxidation in sodium environment is considered to be responsible for the delayed crack initiation and consequent increase in fatigue life. Comparison with RCC-MR code shows that the design curve based on air tests is conservative.

Simulation of the EBR-II Loss-of-Flow Tests Using the MARS Code

As part of the development of a safety analysis methodology for a liquid-metal reactor (LMR) in Korea, the Multidimensional Analysis of Reactor Safety (MARS) code was selected as a system transient safety analysis code. The Korea Atomic Energy Research Institute developed the MARS code to analyze safety and thermal-hydraulic phenomena related to a two-phase flow in the transients of water reactors a decade ago. The addition of thermal-hydraulic models related to liquid metal as a coolant and reactivity feedback models associated with the kinetics calculation of an LMR core is required for the application of the MARS to the transients of an LMR design. A table for various properties of liquid sodium, several heat transfer coefficients according to flow regimes and geometries, and the models for a pressure drop due to the wire spacers of the LMR core were newly implemented. The improved MARS code was verified through the analysis of three shutdown heat removal tests (SHRT)-17, -39, and -45 conducted in the Experimental Breeder Reactor (EBR)-II reactor. The SHRT-17 test involved a simultaneous loss of electrical power to all pumps and a reactor scram from 100% power and flow. Thus, the test simulated a thermal-hydraulic transition from a forced convection to the totally passive decay heat removal due to a natural circulation. SHRT-39 and SHRT-45 are loss-flow tests without a reactor scram. However, the pump coastdown periods and initial states of the plant are different from each other. Simulated results for the flow rate and temperature for an instrumented subassembly agree well with the experimental data.

Development, computer simulation and performance testing in sodium of an eddy current flowmeter

Abstract Sodium is used as a coolant in Liquid Metal Fast Breeder Reactor (LMFBR). Sodium flow measurement is of prime importance both from the operational and safety aspects of a fast reactor. Various types of flowmeters namely permanent magnet, saddle type and eddy current flowmeters are used in FBRs. From the safety point of view flow through the core should be assured under all operating conditions. This requires a flow sensor which can withstand the high temperature sodium environment and can meet the dimensional constraints and be amenable to maintenance. Eddy current flowmeter (ECFM) is one such device which meets these requirements. It is meant for measuring flow in PFBR primary pump and also at the outlets of the fuel sub-assemblies to detect flow blockage. A simulation model of ECFM was made and output of ECFM was predicted for various flowrates and temperatures. The simulation model was validated by testing in a sodium loop. This paper deals with the design, simulation and tests conducted in sodium for the eddy current flowmeter for use in the Prototype Fast Breeder Reactor (PFBR).

Development of one-dimensional computer code DESOPT for thermal hydraulic design of sodium-heated once through steam generators

Once-through Steam Generator (SG) is a critical component of Liquid Metal Fast Breeder Reactor (LMFBR) plant. It is a counter current heat exchanger, in which heat is transferred from the hot sodium flowing on the shell side to water/steam in tube side. High pressure subcooled water enters the SG tube from bottom, gets heated up to saturation, goes through nucleate boiling, dry out and post dry out heat transfer, getting converted to saturated steam and finally gets superheated. For this the process design needs to be carried out accurately. A computer code DESOPT has been developed for the process design of straight vertical, serpentine and helical geometries and validated against reported designs in literature. Recently a test facility to test a 5.5 MWt sodium heated steam generator has been commissioned. The predictions of the code have been compared with the measurements and found satisfactory. This paper brings out different heat transfer mechanisms in SG and describes the one-dimensional code, its validation based on literature and in-house tests and presents the results of comparison between predicted and actual operation at different part loads.

A singular integro-differential equation model for dryout in LMFBR boiler tubes

A 2D steady model for the annular two-phase flow of water and steam in the steam-generating boiler pipes of a liquid metal fast breeder reactor is proposed. The model is based on thin-layer lubrication theory and thin aerofoil theory. The exchange of mass between the vapour core and the liquid film due to evaporation of the liquid film is accounted for using some simple thermodynamics models, and the resultant change of phase is modelled by proposing a suitable Stefan problem. Appropriate boundary conditions for the flow are discussed. The resulting non-linear singular integro-differential equation for the shape of the liquid film free surface is solved both asymptotically and numerically (using some regularization techniques). Predictions for the length to the dryout point from the entry of the annular regime are made. The influence of both the traction τ provided by the fast-flowing vapour core on the liquid layer and the mass transfer parameter η on the dryout length is investigated.

Conceptual evaluation of fluid–structure interaction effects coupled to a seismic event in an innovative liquid metal nuclear reactor

The seismic response analysis of such liquid storage systems, especially liquid metal reactors, as for example the eXperimental Accelerator Driven System (XADS), was examined taking into account mainly the coupling effects of the fluid–structure interaction and their influence on its relevant internal systems and components. Therefore this paper deals with the structural analyses of the seismically induced hydrodynamic responses, in the event of a safe shutdown earthquake (SSE), and the free oscillation (known as sloshing waves) of a metal liquid coolant as well as the dynamic buckling effects on involved structures. To the mentioned purpose the interaction and coupling effects among the main reactor vessel structures and the primary coolant response were investigated by means of a numerical evaluation (with a qualified finite element code) because of the lack of analytical linear theories that in any case are not adequate to describe all the complex phenomena related to the seismic loading. For the numerical modelling procedure, 3D finite element models were set up to analyse the propagation of seismic waves as well as its derived structural effects, such as the fluid steep waves motion, the local buckling bulges, etc., taking into account the geometrical and material nonlinearities of the RPV and the considered simplified internals. The obtained numerical results in terms of stress intensity and of the capability of the structures to resist relevant seismic loads are, thus, presented and discussed. Moreover the performed analyses allowed to highlight the structures mostly affected by the assumed loading conditions in order to achieve data useful for an upgrading of the design geometry, if any, for the considered reactor.

DEVELOPMENT OF THE MATRA-LMR-FB FOR FLOW BLOCKAGE ANALYSIS IN A LMR

The Multichannel Analyzer for Transient and steady-state in Rod Array - Liquid Metal Reactor for Flow Blockage analysis (MATRA-LMR-FB) code for the analysis of a subchannel blockage has been developed and evaluated through several experiments. The current version of the code is improved here by the implementation of a distributed resistance model which accurately considers the effect of flow resistance on wire spacers, by the addition of a turbulent mixing model, and by the application of a hybrid scheme for low flow regions. Validation calculations for the MATRA-LMR-FB code were performed for Oak Ridge National Laboratory (ORNL) 19-pin tests with wire spacers and Karlsruhe 169-pin tests with grid spacers. The analysis of the ORNL 19-pin tests conducted using the code reveals that the code has sufficient predictive accuracy, within a range of 5℃, for the experimental data with a blockage. As for the results of the analyses, the standard deviation for the Karlsruhe 169-pin tests, 0.316, was larger than the standard deviation for the ORNL 19-pin tests, 0.047.