Pool-type Fast Breeder Reactor Research Articles

The numerical simulation of the flow distribution and flow-induced vibration analysis for intermediate heat exchanger in a pool-type fast breeder reactor

Intermediate heat exchanger in a pool-type fast breeder reactor transfers heat from the primary sodium to secondary sodium. A dual cell coupling porous media model is used to simulate flow field in both primary and secondary side at the same time. The velocity distribution in the inlet and outlet are also studied in order to find the velocity change of inlet and outlet. Flow-induced vibration is investigated. On the top of primary side outlet, it is easy to occur flow-induced vibration, support plates or baffles need to be installed here to reduce the radial velocity to avoid flow-induced vibration.

Thermo-mechanical behaviour of primary system components of future FBR during crash Cooling: A numerical simulation

In a pool-type fast breeder reactor, the shutdown by SCRAM involves crash cooling of primary sodium at the rate of 480 K/h. This high cooling rate can develop high thermal stresses at the thicker junctions that are immersed in sodium, due to a lower Fourier number and a higher Biot number, which results in higher Fitz number. Thermo-mechanical analysis of various critical locations within sodium with the transient loading, and parametric studies with varying cooling rates and thicknesses has been carried out. Due to the drop in sodium temperature during crash cooling, there is a fall in the sodium level, resulting from the volumetric contraction of sodium. Thus the locations of the hot pool components that are at the free level interface of sodium and argon experience an abrupt shift in the ambience, i.e., from sodium to argon cover gas. A numerical model has been developed to simulate the complex loading condition using CAST3M code. The model has been validated with theoretical equations and the overall heat balance equations. Transient analysis of all the hot pool components has been performed to estimate the structural damage due to thermo-mechanical behavior during crash cooling. Based on this study, it is concluded that for the future design of fast reactors with 60 years of design life, controlled cooling is not essential from the structural damage point of view. The recommendation simplifies the design of decay heat removal system.

Significance of coast down time on safety and availability of a pool type fast breeder reactor

Plant dynamic investigation towards establishing the influence of flow coast down time of primary and secondary sodium systems on safety and availability of plant has been carried out based on one dimensional analysis. From safety considerations, a minimum flow coast down time for primary sodium circuit is essential to be provided to limit the consequences of loss of flow event within allowable limits. Apart from safety benefits, large primary coast down time also improves plant availability by the elimination of reactor SCRAM during short term power failure events. Threshold values of SCRAM parameters also need optimization. By suitably selecting the threshold values for SCRAM parameters, significant reduction in the inertia of pumping systems can be derived to obtain desirable results on plant availability. With the optimization of threshold values and primary flow coast down behaviour equivalent to a halving time of 8s, there is a possibility to eliminate reactor SCRAM during short term power failure events extending up to 0.75s duration. Benefits of secondary flow halving on reducing transient thermal loading on components have also been investigated and mixed effects have been observed.

Computation of natural convection test at Phenix reactor using the NETFLOW++ code

The present paper describes modeling and analyses of a natural convection test of the pool-type fast breeder reactor (FBR) Phenix. The natural convection test was carried out as one of the end-of-life tests of the Phenix reactor. Objective of the present study is to assess the applicability of the NETFLOW++ code which has been verified thus far using various water facilities and validated using the plant data of the loop-type FBR “Monju” and the loop-type experimental fast reactor “Joyo”. The Phenix primary heat transport system is modeled based on the benchmark documents available from IAEA. The computation model consists of only the primary heat transport system with boundary conditions on the secondary-side of intermediate heat exchangers (IHX). The coolant temperature at the primary pump inlet, the primary coolant temperature at the IHX inlet and outlet, the secondary coolant temperatures and other parameters are calculated by the code where the heat transfer between the hot and cold pools is explicitly taken into account. A model including the secondary and tertiary systems was prepared, and the computed results also agree well with the measured data in general.

Development of a porous body model for decay heat removal studies in a pool type sodium cooled fast reactor

In this paper, the primary heat transport circuit of a pool type fast breeder reactor has been analyzed to demonstrate its decay heat removal capability without exceeding design safety limits even under complete power failure condition leading to natural convection flow in the system. A three dimensional model of large sodium pool (~12 m diameter) and complex inter-wrapper flow path through narrow gaps (3.4 mm) between a large number of heat generating subassemblies is simplified to a semi-3D porous body model which is computationally less intensive. The proposed semi-3D model has been validated against the detailed three dimensional model. The analysis has been carried out using the general purpose CFD code STAR-CD. The predicted results demonstrate that the decay heat removal system is conducive for natural convection and can remove the heat, keeping all the important temperatures within design safety limits.

Investigation on buckling of FBR vessels under seismic loadings with fluid structure interactions

The critical thin walled shell structures in the reactor assembly of a pool type fast breeder reactor (FBR) are the main vessel, inner vessel and thermal baffles. On these structures, the seismic events impose major forces by developing high dynamic pressures, thereby causing a concern on structural integrity due to buckling. An integrated analysis for determining realistic forces and critical buckling loads at any instant during the seismic event has been carried out for the vessels of a typical 500 MWe pool type fast breeder reactor. The dynamic forces including pressure distributions generated on the vessel surfaces are extracted from the seismic analysis carried out for the reactor assembly. The seismic forces thus computed from axisymmetric analysis are transmitted appropriately to 3D shell geometries for the buckling analysis. In view of high computational time needed for carrying out buckling analysis at every time increment, the elastoplastic analysis is carried out only at a few critical time steps which are identified based on strain energies that are associated with the shear and compressive stresses developed at the portions of the vessels prone to buckle. The shear buckling of main vessel straight portion and buckling of toroidal portion of inner vessel and thermal baffles are found to be important. The possible randomness of support excitation time histories is accounted for by compressing as well as expanding the nominal time histories by 10%. Buckling strength reduction factors due to the initial geometrical imperfections are adopted from the literature. The inner vessel is found to be the most critical component which buckles under seismic forces induced by a safe shutdown earthquake with a load multiplier of 1.52, which is higher than the minimum factor of safety of 1.3 required as per the design code RCC-MR [RCC-MR: edition, 2002. Design and construction rules for mechanical components for FBR nuclear islands, vol. 1, section I, subsection B. AFCEN, Paris, in press] for service level D conditions.

Thermal hydraulic investigations of intermediate heat exchanger in a pool-type fast breeder reactor

Intermediate heat exchanger (IHX) in a pool-type liquid metal cooled fast breeder reactor is an important heat exchanging component as it forms an intermediate boundary between the radioactive primary sodium in the pool and the non-radioactive secondary sodium in the steam generator (SG). The thermal loads during steady state and transient conditions impose thermal stresses on the heat exchanger tubes and on the shells which hold the tube bundle. Estimation of these thermal loads and achieving uniform temperature distribution in the tubes and shells by having uniform flow distributions are the major tasks of thermal hydraulic investigations of IHX. Through multi-dimensional thermal hydraulic investigations performed using commercially available computer codes such as PHOENICS, the flow and temperature distributions in the tubes and shells and in its secondary sodium inlet and outlet headers are obtained with and with out provisions of flow distribution devices. The effectiveness of these devices in achieving acceptably uniform flow and temperature distributions has been assessed and thermal loads on the tubes and shells for thermo mechanical analysis of the IHX have been defined. The predictions of the computational studies have been validated against simulated experiments.

Effects of nuclear island connected buildings on seismic behaviour of reactor internals in a pool type fast breeder reactor

The seismic analysis of reactor assembly housing the primary circuit of a typical 500 MWe capacity pool type fast breeder reactor (PFBR) is reported. The reactor assembly is supported on the reactor vault within the nuclear island connected buildings (NICB). The seismic responses, viz. critical displacements, sloshing heights, stresses and strain energy values in the vessels are determined for the reactor assembly by detailed finite element analysis including the fluid–structure interaction and sloshing effects. Analysis is carried out to quantify the effects of inter-connection of the reactor vault with the adjacent buildings under the assumptions that the reactor vault along with reactor assembly is: (1) an isolated structural system from the adjacent buildings within reactor containment building (RCB) and (2) connected with the adjacent civil structures through floor slabs. Analysis indicates that, by inter-connecting the vault with the NICB, there are overall increases of all the governing parameters which decide the seismic design criteria. The significant effects are increases of: (1) radial and axial displacements of core top and absorber rods and vertical accelerations of core subassemblies which are of concern to reactor safety, (2) primary membrane stress intensities for the inner vessel and (3) strain energies developed at the critical portions which can enhance the buckling risks of main vessel, inner vessel and thermal baffles. Hence, it is preferable to isolate the reactor vault, directly constructing from the base raft without inter-connecting it with the NICB, from the seismic loading considerations.

Assessment of Life of Inner Vessel for a Pool-Type Fast Breeder Reactor Based on Inelastic and Viscoplastic Deformations

Towards assessing the creep-fatigue damage for the inner vessel of a pool-type fast reactor, 3-D elastic and axisymmetric inelastic analysis with classical kinematic hardening model, ORNL material model and more realistic CHABOCHE visco-plastic model have been performed. The material properties of stainless steel type 316 LN at various temperatures are used in the analysis. Using the results of this analysis, the creep-fatigue damage and ratcheting are assessed as per the rules of RCC-MR and results have been compared for the various constitutive models used. Inelastic analysis has been performed only at highly stressed local regions. Computations with the ORNL material model is found to be satisfactory in terms of conservatism and computer cost involved, and hence recommended for use in the design stage. Further, conservatism in estimating the creep fatigue-damage of the component associated with the elastic route of RCC-MR has been quantified. The inelastic and viscoplastic analysis tool has been validated using experimental benchmark problems.

Passive Safety Features of a Bottom-Supported Fast Breeder Reactor Vessel

A system dynamics analysis is applied to a pool-type fast breeder reactor to examine the influence of a bottom-supported reactor vessel (BSRV) design on anticipated transient without scram (ATWS) events such as an unprotected loss of flow (ULOF), an unprotected loss of heat sink (ULOHS), and an unprotected transient overpower (UTOP) by using the ARGO safety analysis code. The BSRV enhances negative feedback because of the differential displacement between the core and the control rod as compared with a top-supported reactor vessel. In particular, the BSRV has the potential, especially in a mixed-oxide-fueled core, to mitigate the design requirements to prevent boiling of the coolant during an ULOF and ULOHS through the elongation of the primary flow coastdown and enhancement of the axial expansion of the control rod drive line. In the metallic-fueled core, the effects of the BSRV on the ATWS events are diminished by the limitation of the sodium temperature increase.

Study of Thermal-Hydraulic Characteristics during Decay Heat Removal in a Pool-Type Fast Breeder Reactor

This paper reports that the thermal-hydraulic characteristics of a pool-type fast breeder reactor during decay heat removal are studied experimentally with a one-eighth-scaled three-dimensional reactor model using water as the working fluid. The model is designed to simulate reactor operations from full power to scram transient and decay heat removal conditions as well as two types of decay heat removal systems, a direct reactor auxiliary cooling system (DRACS) and a primary reactor auxiliary cooling system (PRACS). The transient thermal-hydraulic performance of the DRACS is compared with that of the PRACS for the water test results. The results are used to evaluate the accuracy of the multidimensional analysis codes AQUA and THERVIS. The analytical results are generally in good agreement with the measurements.

Static and dynamic buckling of large thin shells: Design procedure, computation tools, physical understanding of the mechanisms

During the last ten years, the French Research Institute for Nuclear Energy (Commissariat à l'Energie Atomique) achieved many theoretical as well as experimental studies for designing the first large size pool type fast breeder reactor. Many of the sensitive parts of this reactor are thin shells subjected to high temperatures and loads. Special care has been given to buckling, because it often governs design. Most of the thin shells structures of the French breeder reactor are axisymmetric. However, imperfections have to be accounted for. In order to keep the advantage of an axisymmetric analysis (low computational costs), a special element has been implemented and used with considerable success in the recent years. This element (COMU) is described in the first chapter, its main features are: • — either non axisymmetric imperfection or non axisymmetric load, • — large displacement, • — non linear material behaviour, • — computational costs about ten times cheaper than the equivalent three dimensional analysis. This paper based on a careful comparison between experimental and computational results, obtained with the COMU element, will analyse three problems: • — First: design procedure against buckling of thin shells structures subjected to primary loads. • — Second: static post buckling. • — Third: buckling under seismic loads.

Report from the I.W.G.F.R. meeting Bologna, Italy (October 8–10, 1984)

The present paper describes modeling and analyses of a natural convection test of the pool-type fast breeder reactor (FBR) Phenix. The natural convection test was carried out as one of the end-of-life tests of the Phenix reactor. Objective of the present study is to assess the applicability of the NETFLOW++ code which has been verified thus far using various water facilities and validated using the plant data of the loop-type FBR “Monju” and the loop-type experimental fast reactor “Joyo”. The Phenix primary heat transport system is modeled based on the benchmark documents available from IAEA. The computation model consists of only the primary heat transport system with boundary conditions on the secondary-side of intermediate heat exchangers (IHX). The coolant temperature at the primary pump inlet, the primary coolant temperature at the IHX inlet and outlet, the secondary coolant temperatures and other parameters are calculated by the code where the heat transfer between the hot and cold pools is explicitly taken into account. A model including the secondary and tertiary systems was prepared, and the computed results also agree well with the measured data in general.

Thermal-Hydraulic Characteristics in a Liquid-Metal Fast Breeder Reactor Hot Plenum

The most important problem in the thermal-hydraulic designs of the pool-type fast breeder reactor is to estimate the thermal conditions affecting the vessel and/or internal structures during both steady and transient operations. The severity of these conditions in the Japanese pool-type reactor, which will be reinforced and equipped with special devices for seismic demands, is apt to be much greater than for other countries. Water tests and thermal-hydraulic analyses have been performed to study such conditions. The effects of the elevations of upper internal structure discharge and intermediate heat exchanger intakes on flow patterns, free surface disturbance, and thermal stratification in the hot plenum have been estimated. From the results of the experiments, suitable elevations could be recommended by comparing some thermal-hydraulic characteristics. The calculations agreed well with the experimental results for the steady-state flow patterns and thermal transients, with the exception of thermal stratification.