Subchannel analysis of supercritical light water-cooled fast reactor assembly

Subchannel analyses have been carried out for supercritical water-cooled fast reactor fuel assembly. Peak cladding surface temperature difference arising from subchannel heterogeneities have been calculated by using the improved subchannel analysis code STARS and was evaluated to be about 18.5 °C. Several suggestions have been also made for reducing the PCST difference arising from channel heterogeneity. Influences of local power peaking on deflection of cladding surface temperature are explained with pin power distribution taken from core depletion calculation in this paper. Maximum cladding surface temperature at nominal condition is evaluated to be 645.3 °C over the cycle. Statistical thermal design uncertainty associated with PCST calculation is evaluated by Monte-Carlo sampling technique combined with subchannel analysis code. Maximum statistical design uncertainty of PCST is calculated to be 31 °C and is in a good agreement with that from RTDP method. Influence of downward flow in seed region on system sensitivity is investigated by improved Monte-Carlo thermal design procedure. Limiting thermal condition of MCST is 681 °C (650 °C of nominal + 31 °C) within 95/95 limit for SWFR.     

Core design of a high-temperature fast reactor cooled by supercritical light water

A high-temperature large fast reactor cooled by supercritical water (SCFR-H) is designed for assessing its technical feasibility and potential economical improvement. The coolant system is once-through, direct cycle where whole core coolant flows to the turbine. The goal is to achieve the high coolant outlet temperature over 500°C. We study the reactors with blankets cooled by ascending and descending flow. SCFR-H adopts a radial heterogeneous core with zirconium-hydride layers between the driver core and the blankets for making coolant void reactivity negative. The coolant outlet temperature of the core with blankets cooled by ascending flow is low, 467°C. The reasons are as follows: (1) the power swing due to the accumulation of fissile material in the inner blankets with burn-up, and (2) local power peak in the assemblies due to the zirconium-hydride layers. The difference in the outlet coolant temperature is more enhanced than the low temperature core where outlet temperature is approximately 400°C. The reason is that the coolant temperature is more sensitive to the enthalpy change than near the pseudo critical temperature, 385°C at 25 MPa. Thus, we design the core with blankets cooled by descending flow to obtain high coolant outlet temperature. The coolant outlet temperature becomes 537°C, which is 70°C higher than that of the core with ascending blanket flow. The thermal efficiency is improved from 43.2 to 44.6%. The coolant mass flow rate per electric power decreases by 14%. This will reduce the size of the balance of plant (BOP) system. The power of the reactor is high (1565 MWe) and the void reactivity is negative.     

European supercritical water cooled reactor

The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 °C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers in the technologies of light water reactors. More than 20 bachelor or master theses and more than 10 doctoral theses on HPLWR technologies have been submitted at partner organizations of this consortium since the start of this project.     

Numerical studies of nuclear traveling waves in a supercritical water cooled fast reactor

An axial fuel shuffling strategy is proposed based on the mechanism of the nuclear fission traveling wave and implemented numerically in the calculation for a supercritical water cooled fast reactor (SCWFR). The ERANOS code is adopted to perform the neutronics and burn-up calculations, and the calculation scheme for axial fuel shuffling and coolant density coupling are set up. The parametric studies of a typical PWR with Th-U and U-Pu (²³⁵U instead of ²³⁹Pu) conversions by burn-up and k_eff calculations indicate that the breeding effects only exist in configurations with very low water content and the conversion or breeding becomes worse as the initial enrichment is increasing. The shuffling calculations for the 1-D SCWFR model described in this paper brought about some interesting results for a certain range of water content. The results indicate that the non-enriched fresh fuel is not possible for both Th-U and U-Pu conversions. As could be expected due to the η-values of the main fissile isotopes ²³³U and (²³⁵U, ²³⁹Pu), respectively, the Th-U conversion needs a lower enrichment, and results in a slightly higher burn-up than the U-Pu conversion. The asymptotic power density distribution of the Th-U conversion is broader and lower than that of the U-Pu conversion. By reducing the water volume fraction, an increased burn-up can be achieved with correspondingly reduced fuel shuffling speed and reduced initial enrichment. Furthermore, the steady state calculations for the asymptotic state show that the Th-U conversion is superior to the U-Pu one concerning SCWFR safety aspects, where the absolute value of the Doppler constant is larger and the coolant feedback is negative for the Th-U conversion, while the coolant feedback is positive for the U-Pu one.     

Thermal hydraulic analysis of supercritical water reactor cooled by TiO2 nanofluid

Heat transfer study of nanofluids as coolant in SCWRs core has been performed at Helwan University. A thermal hydraulic code has been produced to study the effect of TiO₂ nanofluid water based as a coolant with comparison with pure water as a coolant. Various volume fractions of nanoparticles TiO₂ (2, 6 and 10%) were used in order to investigate its effects on reactor thermalhydraulic characteristics. Based on Parameters of a SCW Canadian Deuterium Uranium nuclear reactor (CANDU), the fuel assembly was modeled to study the effect of nanoparticles volume fraction on thermos-physical properties of basic fluid and the temperature distribution of fuel, cladding surface and coolant in axial direction. The theoretical results showed that the density, viscosity and thermal conductivity of the coolant increases with the increase of nanoparticles volume fraction, contrasting to specific heat, which decreases with the increase in nanoparticles volume fraction.     

Design of an indirect-cycle fast breeder reactor cooled by supercritical steam

The neutronic feasibility of a large-sized FBR cooled by supercritical steam is assessed for finding the way to reduce the costs of FBR plants. A negative coolant void reactivity is realized without much deterioration in breeding capability, by our novel concept of inserting thin zirconium-hydride layers between the seed and blanket of the radially heterogeneous core. The gross electric power is 1040 MWe. The estimated equilibrium compound system doubling time is 16 years. The discharge burnup is 78.7 GWd/T and the refueling period is 1 year with a 77% load factor. Compared with the conventional steam cooled FBR, the pumping power fraction is reduced due to the high coolant density of supercritical water. Loeffler boilers for generating steam are not neccessary. The reactor system is simplified and similar to a PWR. The thermal efficiency is 39.6%, improved 15% relatively from PWR's. The pressure vessel is 32.5 cm thick.     

Concept of a pressurized water reactor cooled with supercritical water in the primary loop

A novel concept of a pressurized water reactor with a primary loop cooled with supercritical water is introduced and analyzed in this work. A steam cycle analysis has been performed to illustrate the advantages of such a nuclear power plant with respect to specific power and thermal efficiency. Moreover, a reactor pressure vessel concept including its internals and a suitable core and fuel assembly design are presented overcoming the problems, which arise due to the high heat up of the coolant and the density change involved with it. The core power and coolant density distributions are predicted with coupled neutronic and thermal-hydraulic analyses. The method features the definition of inlet orifices for coolant mass flow adjustment within the core as well as an additional tool for the interpolation of local pin power data. The latter one has been used for a successive sub-channel analysis of the hottest fuel assembly of the core, which provides a more detailed spatial resolution and thus predicts peak cladding temperatures, the maximum linear pin power of fuel pins, and maximum fuel temperatures. It can be shown that maximum temperatures of claddings and fuel are well below the material limits. Thanks to an average core exit temperature below the pseudo-critical temperature, the core concept leaves enough margin for additional uncertainties and allowances for operation.     

Plutonium breeding of light water cooled fast reactors

Light water cooled fast reactor with new fuel assemblies (FA) has been studied for high breeding of fissile plutonium. It achieves fissile plutonium surviving ratio (FPSR) of 1.342 (discharge/loading), 1.013 end and beginning of equilibrium cycle (EOEC/BOEC), and compound system doubling time (CSDT) of 95.9 years at the average coolant density of pressurized water reactor (PWR). It is further improved for reduced moderation boiling water reactor (BWR) (RMWR) coolant density. Fissile plutonium surviving ratio reaches 1.397 (discharge/loading), 1.030 (EOEC/BOEC) and CSDT is 37 years. The present study has shown the possibility of breeding at the PWR coolant density and meeting the growth rate of energy demand of advanced countries at the RMWR and Super FR coolant density for the first time. The new FA consist of closely packed fuel rods. The integrity of welding of fuel rods at the top and bottom ends is maintained as the conventional fuel rods. The coolant to fuel volume fraction is reduced to 0.085, one-sixth of that of RMWR. The volume fraction remains unchanged with the diameter of the fuel rod. The thermal hydraulic design of the cores remains for the future study.     

Thermo-mechanical analysis of supercritical pressure light water-cooled fast reactor fuel rod by FEMAXI-6 code

Thermo-mechanical behaviors of supercritical pressure light water cooled fast reactor (SWFR) fuel rod and cladding have been investigated by FEMAXI-6 (Ver.1) code with high enriched MOX fuel at elevated operating condition of high coolant system pressure (25 MPa) and high temperature (500 °C in core average outlet temperature). Fuel rod failure modes and associated fuel rod design criteria that is expected to be limiting in SWFR operating condition have been investigated in this fuel rod design study. Fuel centerline temperature is evaluated to be 1853 °C and fission gas release fraction is about 45% including helium production. Cumulative damage fraction is evaluated by linear life fraction rule with time-to-rupture correlation of advanced austenitic stainless steel. In a viewpoint of mechanical strength of fuel cladding against creep rupture and cladding collapse at high operation temperature, currently available stainless steels or being developed has a potential for application to SWFR. Admissible design range in terms of initial gas plenum pressure and its volume ratio are suggested for fuel rod design The stress ranges suggested by this study could be used as a preliminary target value of cladding material development for SWFR application.     

Development of natural circulation analysis methods for a sodium cooled fast reactor

A natural circulation evaluation methodology has been developed to insure safety of a sodium cooled fast reactor (SFR) of 1500 MWe adopting a natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can be applied to safety evaluation for SFR licensing taking into account the temperature flattening effect due to buoyancy force in the core, and a three-dimensional fluid flow analysis which can evaluate thermal-hydraulics for local convection and thermal stratification in the primary system and DHRSs. The one-dimensional safety analysis method and the three-dimensional fluid flow analysis method have been validated using the test results of a water test apparatus and a sodium test loop for some typical transient events selected from the design basis events of the SFR. Finally, it has been confirmed that a good agreement between the test results and analysis results has been obtained, and reliability of each method has been demonstrated.     

Safety analysis of a supercritical-pressure water-cooled fast reactor under supercritical pressure

The supercritical-pressure water-cooled fast reactor (SWFR) is a fast spectrum supercritical water-cooled reactor (SCWR) studied by the University of Tokyo. The SWFR is designed as a two-pass core with an outlet temperature 500 °C. The SWFR has fuel channels cooled by downward flow, higher power density, and smaller coolant density reactivity feedback compared with Super LWR. This paper describes the safety analyses of abnormal events for the SWFR. SPRAT-F code is used for the safety analysis at supercritical pressure considering the downward flow cooled seed fuel channel. This code is based on a 1-D node junction model with point kinetics and decay heat calculations. Flow redistribution among parallel paths is calculated by pressure-loss balance and momentum conservation. The initiating events are selected from those of LWRs. For the safety analysis, nine abnormal transients and four accidents are selected with considering types of abnormality. By the numerical analyses, it was found that the loss of flow events can be mitigated by the “water source” effect of the downward flow blanket channels in the abnormal transients and accidents. All the abnormal events satisfy the criteria with margin.     

Gas cooled fast reactor research in Europe

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

Estimation of the amount of surface contamination of a water cooled nuclear reactor by cooling water analysis

Calculations, based upon on-the-spot measurements, were performed to estimate the contamination of NPP primary circuit and spent fuel storage pool solid surfaces via the composition of the cooling water in connection with a non-nuclear incident in the Paks NPP. Thirty partially burnt-up fuel element bundles were damaged during a cleaning process, an incident which resulted in the presence of fission products in the cooling water of the cleaning tank (CT) situated in a separate pool (P1). Since this medium was in contact for an extended period of time with undamaged fuel elements to be used later and also with other structural materials of the spent fuel storage pool (SP), it was imperative to assess the surface contamination of these latter ones with a particular view to the amount of fission material. In want of direct methods, one was restricted to indirect information which rested mainly on the chemical and radiochemical data of the cooling water. It was found that (i) the most important contaminants were uranium, plutonium, cesium and cerium; (ii) after the isolation of P1 and SP and an extended period of filtering the only important contaminants were uranium and plutonium; (iii) the surface contamination of the primary circuit (PC) was much lower than that of either SP or P1; (iv) some 99% of the contamination was removed from the water by the end of the filtering process.     

Theoretical and experimental thermal-hydraulic analysis of a new sodium cooled fast breeder reactor core

This paper presents the experimental and theoretical results of the thermal-hydraulic design of a new fast breeder reactor core concept. The main feature of this concept is the omission of fuel element cans.The hydraulic function of these fuel element cans is substituted by a winding flow path through the radial blanket and a ring chamber without tubes.A computer code based on the quasi-continuum-theory and especially adapted to the features of the new core concept is developed for theoretical investigations. The pressure drop of the rod bundles is specified by a resistance tensor.The experimental investigations are realized in a test facility, where sodium is simulated by water. Pressures and velocities are measured.Theoretical and experimental results show good agreement. The aim of flattening of the coolant outlet temperature distribution can be reached with satisfying accuracy.     

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

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

Numerical analysis of heat transfer in supercritical water cooled flow channels

Research activities are ongoing worldwide to develop nuclear power plants with a supercritical water cooled reactor (SCWR) with the purpose to achieve a high thermal efficiency and to improve their economical competitiveness. However, there is still a big deficiency in understanding and prediction of heat transfer in supercritical fluids. In this paper, heat transfer of supercritical water has been investigated in various flow channels using the computational fluid dynamics (CFD) code CFX-5.6 to provide basic knowledge of the heat transfer behaviour and to gather the first experience in the application of CFD codes to heat transfer in supercritical fluids. Three different flow channels are selected, i.e. circular tubes, the sub-channel of a square-array rod bundle and the sub-channel of a triangular-array rod bundle. The effect of mesh structures, turbulence models, as well as flow channel configurations is analysed. Based on the present results, recommendations are made on the application of turbulence models to the heat transfer of supercritical fluids in various flow channels. A new definition for the onset of heat transfer deterioration is proposed. A strong non-uniformity of heat transfer is observed in sub-channel geometries. This non-uniformity has to be taken into account in the design of fuel assemblies of SCWR.     

Lead-bismuth cooled fast reactor for regional power generation

A condition for adopting low-capacity nuclear power plants in regional power generation is that they be competitive with thermal power plants. However, it is much more difficult to make low-capacity plants competitive than the power-generating units of high-capacity nuclear power plants, because as the capacity of a power source decreases the specific capital investments and power generation costs increase much more rapidly. It is shown that the innovative nuclear power technology of lead-bismuth cooled fast reactors, such as SVBR, based on the experience gained in operating nuclear submarines with chemically inert lead-bismuth coolant, which does not require high pressure in the first loop, satisfies all requirements for low-capacity nuclear power plants for regional power generation.     

Gas cooled fast reactor for generation IV service

The Gas Cooled Fast Reactor (GFR), which is among the Generation IV concepts under evaluation for future deployment, will have to satisfy the Gen IV goals in the area of sustainability, safety and economy. This paper discusses challenges posed by the GFR when striving for the achievement of balance among the above Generation IV goals, and the pros and cons of various design choices. Considering these goals, the currently preferred design direction at MIT is a GFR design using a direct supercritical CO₂ cycle, traditional containment with design pressure of 5 bars, employment of redundant active emergency cooling systems with highly reliable and diverse power supplies, which can also function in the passive mode as a backup at 5 bars containment pressure, and TRU fueled cores using either block-type (TRU-U)C fuel or pin type (TRU-U)C fuel with double cladding or (TRU-U)O₂ fuel vibropacked in a tube-in-duct assembly.     

Transmutation of americium in a medium size sodium cooled fast reactor design

We have performed transient analysis of a medium size sodium cooled reactor loaded with different fractions of americium in the fuel. Unprotected Loss of Flow (ULOF) and Unprotected Transient over Power (UTOP) accidents were simulated in a geometrical model of BN600, using safety parameters obtained with the SERPENT Monte Carlo code.     

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

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