Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

Unprotected loss of flow (ULOF) analysis of metal (U–Pu–6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500 MWe and 1000 MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500 MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000 MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000 MWe design is assured, even with 20% uncertainty on the sensitive parameters.     

Steady state and transient sodium boiling experiments in a 37-pin bundle

As part of the fast breeder reactor safety analysis steady state and transient sodium boiling tests were performed out-of-pile in an electrically heated 37-pin bundle. The steady state boiling experiments served for investigations of the two-phase flow physics and to support the analysis of the transient experiments. The experimental work concentrated on the transient sodium boiling tests which simulated the unprotected loss of flow accident (ULOF) from the start of the flow run down via boiling inception to the onset of dryout. Special emphasis was laid upon the analysis of the transition from the spatial to the mainly one-dimensional growth of the boiling region during the flow transient. The experimental results from both types of tests serve as data basis for computer code validations. A reference test (L22) of the transient experiments was satisfactorily recalculated with a one-dimensional and with a three-dimensional computer programme.     

Startup transient simulation for natural circulation boiling water reactors in PUMA facility

In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs.     

Experimental research on heat transfer to liquid sodium and its incipient boiling wall superheat in an annulus

Liquid sodium is mainly used as a cooling fluid in the liquid metal fast breeder reactor （LMFBR）, whose heat transfer, whether convective heat transfer or boiling heat transfer, is different from that of water. So it is important for both normal and accidental operations of LMFBR to perform experimental research on heat transfer to liquid sodium and its boiling heat transfer. This study deals with heat transfer with high temperature （300-700℃） and low Pe number （20-70） and heat transfer with low temperature （250-270℃） and high Pe number （125-860）, and its incipient boiling wall superheat in an annulus. Research on heat transfer involves theoretical research and experiments on heat transfer to liquid sodium. It also focuses on the theoretical analysis and experimental research on its incipient boiling wall superheat at positive pressure in an annulus. Semiempirical correlations were obtained and they were well coincident with the experimental data.     

A comparative study of unprotected loss of flow accidents in 500 MWe FBR metal cores with PFBR oxide core

Analyses of unprotected loss of flow accidents for 500 MWe U–Pu–6%Zr and U–Pu–10%Zr metal fueled sodium cooled reactors are presented and compared with that of the 500 MWe (U–Pu) MOX Prototype Fast Breeder Reactor (PFBR – under construction in Kalpakkam, India). A flow halving time of 8 s is considered for all the cases. It is found that the results of the metal fuel cases are close to each other. The loss of flow accident is benign for the metal fueled reactors where it is found that sodium coolant boiling is delayed up to 900 s, without credit for safety grade decay heat removal systems. In contrast, the oxide fueled reactor shows much earlier onset of sodium boiling and fuel slumping, leading to near prompt criticality and entry into the disassembly phase. Thus it is concluded that unprotected loss of flow accidents in metal fueled reactors are benign and allow sufficient time for operator action, if safety grade decay heat removal systems are able to remove the decay heat.     

Sodium boiling experiments in a seven-pin bundle: Flow patterns and two-phase pressure drop

Two series of quasi-steady state sodium boiling experiments have been carried out in an electrically heated seven-pin bundle. The power levels (130–170 and 30–40 W/cm²) and other test conditions were selected to correspond to the core and radial breeder zones of a typical LMFBR. The test procedure involved the gradual reduction of mass flow rate through the bundle in a simulation of the consequences of a slowly growing blockage in the lower part of a reactor subassembly. By this means it was possible to study the development of quasi-steady state boiling up to the onset of permanent dryout. The results obtained provide information on flow regimes in the two-phase region, vapour qualities and flow rates at which cooling of the bundle can be effectively maintained, and the ultimate incidence of dryout. A relation for the two-phase pressure drop multiplier obtained from adiabatic pressure drop measurements in this geometry is given and compared with earlier results from single-channel geometry tests.     

The development of techniques for the surveillance of LMFBRs

The development of surveillance techniques of LMFBRs is determined by the interaction of three factors: the specification of requirements, improvements in technique and the physical analysis of the processes involved. The specification of requirements, which sets the structure for the discussion, is mainly concerned with public safety. Two main divisions are identified: those concerned with thermal events in the nuclear core and those concerned directly or indirectly with the mechanical integrity of components. The necessary developments are then discussed in terms of the signal analysis techniques to anticipate various modes of failures. The importance of an adequate understanding of the failure mode is emphasised in optimising the surveillance technique.

Core surveillance may be achieved by monitoring individual sub-assemblies or by monitoring bulk conditions. The important features of sub-assembly monitoring are discussed and the advantages of temperature analysis explained. The specification of the temperature-monitoring systems is identified and the conflicting requirements for the reactor sensor discussed, viz adequate band width as against a robust and reliable construction. A theoretical treatment using Monte Carlo techniques allows a full examination of the choice of method of temperature analysis. This shows that, although a filtered rms value has been the preferred choice for detecting either local blockage or sodium boiling, it may be possible to distinguish the temperature signals of blockages from those of power gradients by an amplitude probability density plot. The advantages of acoustic monitoring using the noise of boiling sodium to detect overheating, leading to core damage, are examined. An important consideration is the thermal-acoustic process of sodium boiling, and evidence is submitted from a range of out-of-pile experiments involving local sub-cooled boiling and bulk boiling in discussing the merits of pulse analysis and power spectral density techniques. An important factor in discriminating background from signal is the extent of cavitation in reactor components. Experiments are mentioned in which pulse techniques have been used to locate boiling sources by spatial correlation. The interpretation of reactor signals requires a detailed knowledge of the transmission of acoustic waves in reactor pools and structures and the effect of gas bubbles. Measurements in PFR and sodium loops have helped to lead to a more quantitative assessment of the sensitivity of the acoustic techniques.

Structural integrity depends on detecting failure modes, particularly those arising from crack propagation. Manufacturing defects or pre-existing cracks may be identified by ultrasonic inspection or by stress-wave emission. On-line monitoring for stress-loaded cracks by a stress-wave emission is seen as intrinsically difficult because of low signal strength and high attenuation but initial experiments have indicated possibilities for detecting stress-corrosion cracking. Mechanical failure from fatigue may be anticipated from a understanding of the vibrational modes of the sodium and its coupling with the structure. A one-eighth scale model of a LMFBR design has recently demonstrated the likely vibrational modes. A major handicap in supervising mechanical operation in sodium systems is the opacity of the sodium. Visualisation techniques of the major parts of the core structure are being developed. An important aspect is the study of the information processing required to present an image easy for the reactor operator to understand. Advances may be made using transform methods to improve object boundaries by modifying the spatial frequencies of the display or record.     

Results of acoustic and temperature noise measurements on sodium cooled electrically heated 7 rod bundle

Assembly cooling deficiency in a LMFBR is one of the most important safety problems for reactor design and operation.

Studies on early detection and diagnosis of local accident by means of noise analysis techniques have been initiated at CNEN. Acoustic and temperature noise measurements have been carried out on a 7 rod bundle during slow power transients up to boiling conditions. The test section, simulating the italian PEC reactor fuel element, was mounted on ENA-2 sodium loop located at the CSN Casaccia.

Acoustic noise spectral analysis up to 32 kHz shows the appearance, in presence of boiling, of power increase at certain frequencies. Power spectra and rms values are updated and recorded every 0.3 sec and show large variations going from single phase to boiling.

Temperature noise spectral analysis shows that the power, between 1 and 50 Hz, increases, in presence of boiling, by a factor bigger than 30. It has been tested the sensitivity of other indicators of the temperature fluctuations, like skewness and flatness, to reveal boiling.     

Sabena: Subassembly boiling evolution numerical analysis

This paper describes the computer code SABENA that has been used in subassembly sodium boiling evolution numerical analysis as a contribution to fast breeder reactor safety analysis. SABENA is a two-fluid model subchannel code system to calculate coolant boiling and two-phase flow in a rod bundle together with external loop characteristics which affects the overall boiling behavior in the bundle section. With the use of relatively simple but reasonable constitutive models, the SABENA code has been applied to and validated against many multi-pin sodium boiling problems. The results have shown excellent agreement with the experiments. The numerical methods and models employed in the code have proven to be robust and efficient in light of the extreme severity of the conditions characterizing low-pressure sodium boiling.     

Flow rundown experiments in a seven pin bundle

A series of experiments has been carried out using an electrically heated seven pin bundle to simulate the conditions under which boiling of the sodium coolant could occur in the event of the loss of power to the circulator pumps of a fast reactor, coincident with a failure of the reactor to trip. Although it was not possible to represent the conditions of the reactor exactly, nor to continue the tests far into dryout, the results nevertheless give valuable qualitative information on the course of boiling development as well as useful quantitative information against which the predictions of computer codes can be checked. In particular, data have been obtained relating to the incidence of superheat, the location and time to dryout of the residual liquid films, the void fraction within the boiling region, and the types of flow regimes which may be expected within different parts of the boiling region at various stages of the transient.     

Experimental investigation of the formation of critical boiling nuclei in superheated sodium

This work deals with the problem of boiling nuclei growth up to the critical size in superheated sodium. The sodium was superheated to a certain degree without initiating the onset of boiling. The lifetime of the superheated state — defined as the waiting time — was measured. The experiments were performed with stagnant sodium in a gas-free apparatus. The sodium temperature and superheat were up to 960°C and 360°C, respectively. Waiting times of up to some hours were measured with large scattering of values. Cold trap purifications influenced the waiting time irreversibly. Natural boiling nuclei were unstable. An artificial boiling nucleus (small cylindrical cavity) was inactive. The existence of many cavities in the test chamber surface was demonstrated by a scanning electron microscope. Experimental results are compared with theoretical predictions from known nucleation models. The dynamic model of the formation of collective heterogeneous boiling nuclei agreed qualitatively.     

Phenomenological investigation of sodium boiling in a SFR core during a postulated ULOF transient with CATHARE 2 system code: a stabilized boiling case

In the framework of the research and development on GEN IV sodium fast reactors (SFRs), the phenomenology of sodium boiling during a postulated unprotected loss of flow (ULOF) transient has been investigated with the CATHARE 2 system code. This study focuses on a stabilized boiling case: in such a regime, no flow redistribution occurs from the subassemblies which have reached the saturation temperature to those that are still single-phase. In this paper, for a subassembly design featuring no restrictive structures above the fuel bundle, a quasi-static approach is first developed to get an upper bound of the reactor core power at boiling onset that would be compatible with the well-known Ledinegg criteria for diphasic flow static equilibrium. Then, dynamics results achieved through simulation with the CATHARE 2 code for a postulated ULOF are presented: boiling is shown to remain stable during the transient for such a core power at boiling onset. Another important outcome of the simulation is the calculation of a dynamic instability, in the form of a two-phase hydrodynamic chugging phenomenon. The predicted phenomenology of this stabilized boiling case should be studied further in order to consider its dependency on the underlying closure laws and to eliminate the possibility of a numerical instability.     

Natural Convection Sodium Boiling Experiments in 37-Pin Bundle Geometry

Decay heat removal capability under boiling condition was studied using an LMFBR fuel subassembly mockup loop. The sodium flow was driven by natural convection through the loop in which was installed a 37-pin bundle heated electrically over a length of 45 cm.

The heat flux furnished by the pins was increased stepwise, upon which the two-phase flow regime changed from bubble to slug flow and then to annular or annular mist flow. Dryout occurred even in slug flow regime, but only momentarily, and permanent dryout was not observed before establichment of annular flow. A suitable criterion for permanent dryout is considered to be 0.5 average exit sodium vapor quality. The results indicated that upon occurrence of sodium boiling, the coolability of fuel subassembly would be maintained by natural convection after reactor shutdown.     

PbBi-cooled direct contact boiling water small reactor

A design concept of PbBi cooled direct contact boiling water small fast reactor (PBWFR) has been formulated with some design parameters identified. Water is injected into hot PbBi above the core, and direct contact boiling takes place in chimneys. Boiling bubbles rise due to buoyancy effects, which works as a lift pump for PbBi circulation. The generated steam passes through separators and dryers for the removal of PbBi droplets, and then flows into turbines for the generation of electricity. The system pressure of 7 MPa is as the same as that of the conventional boiling water reactors (BWRs). The outlet steam is superheated by 10°C to avoid the accumulation of condensate on a PbBi free surface in the reactor vessel. The control rods are inserted from above, which is different from the original concept. This insertion was chosen since the seal of steam at the top of the reactor vessel is technically much easier than the seal of PbBi at the bottom of the reactor vessel. The electric power of 150 MWe may be the maximum which is practically possible as a small reactor with economic competitiveness to conventional LWRs. A two-region core is designed. A decrease in reactivity was estimated to be 1.5%dk/kk′ for 15 years. A fuel assembly has 271 fuel rods with 12.0 mm in diameter and 15.9 mm in pitch in a hexagonal wrapper tube. The design limit of cladding temperature is specified to be 650°C for compatibility of cladding material with PbBi. As a result, the PbBi core outlet temperature becomes 460°C. The PbBi temperature rise in the core is 150°C. The conditions of the secondary coolant steam are as the same as those of conventional BWRs with thermal efficiency of 33%. The core is designed to have the breeding ratio of 1.1 and the refueling interval of 15 years as a reactor with a long-life core. Direct heat exchangers (DHX), reactor vessel air cooling systems (RVACS) and guard vessel are designed.     

Experimental investigation on self-leveling behavior in debris beds

During a hypothetical core-disruptive accident in a sodium-cooled fast breeder reactor, degraded core materials can form debris beds on the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of core material pool. Coolant boiling may lead ultimately to leveling of the debris bed that is crucial to the relocation of molten core and heat-removal capability of the debris bed. In the present study, we elected to use depressurization boiling to simulate an axially increasing void distribution in the debris bed. Bottom-heating boiling was also chosen to confirm that characteristics of the self-leveling process do not depend on the boiling mode. Particle size (between 0.5 and 6 mm), shape (spherical and non-spherical), bed volume (between 5 and 8 l) and density (namely of alumina, zirconia, lead and stainless steel) along with boiling intensity and total volume were taken as experimental parameters to obtain the general characteristics of the self-leveling process. A series of experiments with simulant materials were conducted and analyzed in detail. The good concordance of the transient processes obtained from the different boiling methods sufficiently demonstrates that the present results obtained using the depressurization boiling method exhibit these general self-leveling characteristics. Detailed comparisons of deduced time variations of the inclination angle provides qualitative tendencies based on the experimental parameters considered influential to self-leveling behavior. The rationale behind the definition introduced for equivalent power density is also presented.     

The detection of sodium vapor bubble collapse in a liquid metal fast breeder reactor

Sodium boiling detection utilizing the sound pressure emanated during the collapse of a sodium vapor bubble in a subcooled media is discussed in terms of the sound characteristic, the reactor ambient noise background, transmission loss considerations and performance criteria. Data obtained in several loss of flow experiments on Fast Test Reactor Fuel Elements indicate that the collapse of the sodium vapor bubble depends on the presence of a subcooled structure or sodium. The collapse pressure pulse was observed in all cases to be on the order of a kPa, indicating a soft type of cavitational collapse. Spectral examination of the pulses indicates the response function of the test structure and geometry is important. The sodium boiling observed in these experiments was observed to occur at a low (<50°C) liquid superheat with the rate of occurrence of sodium vapor bubble collapse in the 3 to 30 Hz range. Reactor ambient noise data were found to be due to machinery induced vibrations, flow induced vibrations, and flow noise. These data were further found to be weakly stationary enhancing the possibility of acoustic surveillance of an operating Liquid Metal Fast Breeder Reactor. Based on these noise characteristics and extrapolating the noise measurements from the Fast Flux Test Facility Pump (FFTP), one would expect a signal to noise ratio of up to 20 dB in the absence of transmission loss. The requirement of a low false alarm probability is shown to necessitate post detection analysis of the collapse event sequence and the cross correlation with the second derivative of the neutronic boiling detection signal. Sodium boiling detection using the sounds emitted during sodium vapor bubble collapse are shown to be feasible but a need for in-reactor demonstration is necessary.     

Transformation and fragmentation behavior of molten metal drop in sodium pool

In order to clarify the fragmentation mechanism of a metallic alloy (U–Pu–Zr) fuel on liquid phase formed by metallurgical reactions (liquefaction temperature = 650 °C), which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum (melting point = 660 °C) and sodium mainly under the condition that the boiling of sodium does not occur. When the instantaneous contact interface temperature (T_i) between molten aluminum drop and sodium is lower than the boiling point of sodium (T_c,bp), the molten aluminum drop can be fragmented and the mass median diameter (D_m) of aluminum fragments becomes small with increasing T_i. When T_i is roughly equivalent to or higher than T_c,bp, the fragmentation of aluminum drop is promoted by thermal interaction caused by the boiling of sodium on the surface of the drop. Furthermore, even under the condition that the boiling of sodium does not occur and the solid crust is formed on the surface of the drop, it is confirmed from an analytical evaluation that the thermal fragmentation of molten aluminum drop with solid crust has a potential to be caused by the transient pressurization within the melt confined by the crust. These results indicate the possibility that the metallic alloy fuel on liquid phase formed by the metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt.     

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs.     

Temperature fluctuations: An assessment of their use in the detection of fast reactor coolant blockages

In this paper, the temperature noise technique for the detection of local blockages in fast reactor subassemblies is discussed. The main factors involved in an assessment of the technique are outlined and the experimental and theoretical work that has been carried out at BNL on the various aspects of the problem is described. It is concluded that blockages appreciably smaller than those predicted to produce boiling should be detectable against a background noise level due to subassembly power tilts, on a time scale giving protection against rapidly developing incidents. Further work required to increase confidence in the application of the technique to the reactor is outlined, including measurements in fully representative geometries, data from sodium rigs and further information on reactor background noise levels.     

Fast boiling reactors with the fuel in the form of a fused salt

An attempt is made to classify fast reactors . Versions of a fast reactor with liquid fuel in a boiling coolant - solvent are considered separately. The advantages of this type of reactor and the feasibility of its construction are evaluated; the basic characteristics of the reactor are derived.Institute of Nuclear Research, Warsaw-Zeran, Polish People's Republic. Translated from Atomnaya Énergiya, Vol.22, No.1, pp. 10–13, January, 1967.