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The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop a subchannel void distribution correction model for the two-stage core analysis method using an assembly-based thermal-hydraulics calculation in the core analysis stage. The model assumes two kinds of subchannel void distribution gradients along with the two diagonal lines in the horizontal cross section of a BWR fuel assembly. The model appends and tabulates the difference of the subchannel perturbation condition from the base condition in the lattice physics, and evaluates the tilts within the 2D lattice physics scheme, and couples those results with 3D subchannel analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. The developed model is evaluated using a heterogeneous and a small core problem. The model gives a better power distribution compared with that of the authors’ previous model. As a result, the model can incorporate the subchannel effect into the current two-stage core calculation method.

The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop a core analysis model coupled with subchannel analysis within the two-stage calculation scheme using an assembly-based thermal-hydraulics calculation in the core analysis. The model changes the 2D lattice physics scheme, and couples with 3D subchannel analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. In order to couple with these two analyses, some BWR fuel assembly parameters are assumed and verified. The developed model is evaluated for the heterogeneous problem with and without a control rod. The present model is especially effective for the control rod inserted condition. The present model can incorporate the subchannel effect into the current two-stage core calculation method.

The neutron flux and fission power profiles through a fuel bundle and across a fuel element are important aspects of nuclear fuel analysis in multi-scale/multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs) with advanced fuel bundles. Fuel channels in many existing PHWRs are horizontal. With ageing, pressure tubes creep and fuel bundles in these pressure tubes are eccentrically located, which results in an asymmetric coolant flow distribution between the top and bottom of the fuel bundles. The diametral change of the pressure tube due to creep is not constant along the fuel channel; it reaches a maximum in the vicinity of the maximum neutron flux location. The cross-sectional asymmetric positioning of fuel bundles in a crept pressure tube contributes to an asymmetric power distribution within a ring of fuel elements. Modern reactor physics lattice codes (such as WIMS-AECL 1 ) are capable of predicting the details of power distribution from basic principles. Thermalhydraulics subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution within and across fuel elements (e.g., flux tilt model, different powers in different ring elements, or radial power profiles). In this work, physics and thermalhydraulics codes are applied to quantify the effect of eccentricity of a fuel bundle on power variations across it and within a fuel element, and ultimately on the fuel temperature distribution and fuel centerline temperature, which is one of the indicators of fuel performance under normal operating conditions (NOC). Due to fissioning, Xe and Kr gases are produced, which contribute to the internal gas pressure within an element. The increase in gas pressure may jeopardize the fuel element integrity (sheath rupture). On the other hand, fission gas is produced as a function of neutron flux/power and diffuses to the grain boundaries and fuel-pellet interface as a function of fuel temperature. Therefore, it is important to know the fuel temperatures and distributions within the fuel.

In this study, the use of the CANFLEX fuel bundle in the CANDU6 reactor has been investigated with high-fidelity Monte Carlo simulations in view of major safety parameters. The CANFLEX fuel bundle is designed such that it has a smaller linear power and flatter bundle power distributions than those of the standard design. This results in a significantly lower average fuel temperature. In addition, with slightly larger coolant flow area than the standard design, the CANFLEX bundle is expected to improve fuel temperature coefficient (FTC) and coolant temperature coefficient (CTC) during normal operation. This study presents 2-D neutronic lattice evaluations of these safety parameters of the CANDU6 reactor loaded with CANFLEX fuel bundle against with the standard fuel design. To adequately model the equilibrium CANDU6 core, a number of different 2-D fuel bundle lattices were simulated on a huge number of neutron histories. For a reasonably accurate evaluation of the FTC, thermal motion of 238 U was considered with the Doppler Broadened Rejection Correction (DBRC) scheme. Based on these numerical evaluations, it can reasonably be concluded that the CANFLEX fuel design can possibly enhance safety of the CANDU6 reactor as its PCR is clearly negative for a wide range of reactor power. All depletion calculations were completed using the Monte Carlo Serpent2 code with ENDF/B-VII.0 library.

In the PWR core, the fuel assembly is firmly seated on the lower core plate during operation. However, if the hydraulic force exerted on the fuel assembly by coolant flow is too large and the fuel assembly is lifted-off from the lower core plate, the excessive vibration will cause fuel failure. Therefore, the hydraulic lift-off issue needs to be addressed when the advanced fuel assembly is developed. It has been shown that the advanced annular fuel design with internal cooling allows power uprating up to 50% while the peak temperature of the fuel can be reduced and the MDNBR can be maintained. However, if the coolant condition in the core is kept unchanged, increasing the core power by 50% requires the core flow rate also increase proportionally, which will give rise to the hydraulic lift-off, an important issue to be addressed. In this paper, taking the 17 × 17 solid fuel design as the reference, the hydraulic lift-off issue is investigated for proposed 12 × 12 and 13 × 13 annular fuel designs. Both the steady-state and start-up operating conditions are evaluated. It is found that the hydraulic lift-off indeed is an issue for annular fuel design which requires careful analysis. By comparison, the lift-off forces and hold-down forces required for the externally and internally cooled annular fuels (13 × 13 and 12 × 12 arrays) are several times larger than that of the referenced solid fuel (17 × 17 array). Therefore, the hold-down mechanism for annular fuel needs to be carefully designed.

The present work has been carried out in the framework of the European FP7 project SEARCH, in which the MYRRHA demonstrator reactor is designed to be able to operate both in ADS mode and in critical mode using lead–bismuth eutectic (LBE) as primary coolant. According to the project task definition, the pin failure and fuel dispersion scenarios in severe accidents had to be extensively studied for reactor safety analysis. In this paper, the unprotected severe transients analyses for the MYRRHA–FASTEF critical core were performed using the SIMMER-III code. The aim of the current work is to obtain a deeper understanding of core material redistribution processes after pin damage. Since the fuel has almost the same density as the coolant, its pellets, chunks and particles will essentially be carried by the coolant flow, thus moving upwards out of the core and redistributing into the upper pool region and peripheral structures. Starting the simulations from the steady state configuration, relevant parameters reflect good agreement with the design operational conditions. For the transients, the most severe accident scenario proposed, that may possibly lead to pin failure and furthermore core damage, is the unprotected blockage accident (UBA). The calculation results show that after pin failure, the mobile fuel starts to re-distribute. In the meantime, the reactor stabilizes to shut-down status because of the fuel loss. Our results show that the blockage propagation is impossible thanks to the gap between fuel subassemblies.

Limited results of studies on coolant channel blockage accidents in a prismatic-gas cooled reactor core are available, although coolant channel blockage of the fuel block in a prismatic gas-cooled reactor has been regarded as a highly important phenomenon. In this study, therefore, intensive and extensive thermal analyses were carried out to enhance the understanding and knowledge about the thermal performance of multi-hole type prismatic fuel blocks under coolant channel blockage accidents. A series of thermo-fluid calculations were performed using the CORONA code. In contrast to existing works, a full core model was used to consider an accurate coolant flow redistribution after the blockage in this work. The results of the CORONA calculations show that the single channel blockage accidents of PMR200 are not a safety concern regardless of the location. However, the maximum fuel temperature exceeds the safety limit in the case of more than ∼10 channel blockage at the same region of the fuel column. It was also found that the selection of the computational domain is crucial. A full core analysis is required for accurate results in the case of severe blockage accidents.

A numerical investigation into the effect of a coastdown flow on the early stage cooling of the reactor pool in Korea Advanced Liquid Metal Reactor (KALIMER)-600 during a loss of normal heat sink accident has been carried out. Based on the design values of KALIMER-600, thermal-hydraulic calculations for steady and transient states have been done using the COMMIX-1AR/P code. Coastdown flow effect was evaluated based on a transient analysis of reactors employing various flywheels, which had coastdown flow time (CDT) values ranging from 0 (without a flywheel) to 300 s. The transient analysis has been done from a reactor trip to the onset of an overflow into the DHX support barrel. It was found that the coastdown flow range could be divided into three zones, based on its effect. Among them an excessive core coolant peak temperature and a reversed flow at the core region were observed for a medium coastdown flow range. The medium ranged coastdown flow induces the development of a high density layer near the core exit. This layer contributes to the development of an adverse effect in the core coolant flow, and finally results in increasing the core peak temperature. It was also found that the initiation of heat removal by DHX could be accelerated by the increase of the CDT, although it needs a large flywheel. From this analysis the best CDT is determined to be 25 s.

The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly.

The present work is concerned with a power upgrading study of Tehran Research Reactor (TRR). The upgrading study is aimed at investigating the possibility of raising power of the TRR from the current level of 5 MW th to a higher level without violating the original thermal-hydraulic safety criteria. The existing core, comprising 22 standard fuel elements and five control fuel elements, is used for the analyses. Different reactor thermal powers (5–11 MW) and different core coolant flow rates (500–921 m 3 /h) are considered. It is shown that, for the present core, this goal could be achieved safely by gradually opening the butterfly control valve until the desired coolant flow rate is reached. The TRR power could be upgraded up to around 7.5 MW th with the total power peaking factor maintained at less than or equal to 3.0.

Nuclear vendors and utilities perform numerous simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes, most of which were developed based on one-dimensional lumped parameter models. During the past decade, however, computers, parallel computation methods, and three-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. The use of advanced commercial CFD codes is considered beneficial in the safety analysis and design of NPPs. The present work analyzes the flow distribution in the downcomer and lower plenum of Korean standard nuclear power plants (KSNPs) using STAR-CD. The lower plenum geometry of a PWR is very complicated since there are so many reactor internals, which hinders in CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of a PWR is used. The results give a clear figure about flow fields in the downcomer and lower plenum of a PWR, which is one of major safety concerns.

The radial-reflector subassemblies of the Experimental Breeder Reactor-II (EBR-II) are in the form of large hexagonal-shaped steel blocks inserted into a hexagonal-shaped steel duct with sodium coolant flowing upward in the thin space between the blocks and duct. This geometry combined with the small coolant flow rates and gamma energy deposition rates cause the coolant temperature distributions to be greatly affected by intersubassembly heat transfers. Reported are calculated trends of the coolant temperatures in the six facial intraduct gaps of radial reflector subassemblies surrounding a seven-rowed core and a six-rowed core. Effects of overall changes of reflector region flows and powers are also examined.

BATAN-MPASS code, a general fuel management code for calculating the core equilibrium condition of pebble-bed high temperature reactors with multipass and once-through-then-out (OTTO) fueling schemes, is developed. The code adopts an iterative method to obtain directly both the equilibrium core and critical conditions of the reactor. Two-dimensional, r − z reactor geometry, as well as optional 1-D reactor geometry are available. The code is also equipped with 1- and 2-D thermal-hydraulic routines to assess the coolant flow pattern and fuel temperature distribution in the core.

A new type of inherently-safe BWR plant design has been conceived. A feature of this design is the combination of the fluidized-bed concept and the density-lock mechanism of PIUS. In this design, channels in which spherical fuel elements fluidize hydraulically connect with a cooling water tank through the density-lock mechanism. Hence the fuel elements naturally fall down in the tank in any accidents including LOCA and ATWS, and are kept under subcritical condition and cooled to remove their decay heat. A preliminary fluidized experiment suggested that the fuel elements fluidized excellently at a coolant flow velocity of half of that in BWR and their deformation due to collisions with each other and of them with the channel wall was not concern to their integrity.

In this paper, the results of flow measurements using axially displaced thermocouples in the channel wall will be introduced. The experiments have been performed at a simulated MRT-type fuel assembly located in the research reactor HOR of the Interfaculty Reactor Institute, Delft. It was found that the velocities obtained via temperature noise correlation methods are significantly larger than the area-averaged velocity in the single-phase coolant flow. Model calculations show that the observed phenomenon can be explained by the effects caused by the radial velocity distribution in the channel.

Frequency composition of inherent coolant boiling noise in BWRs is analysed using the equations governing the bubble formations and their detachment from the fuel-pin surface and a parametric study is conducted with respect to core cooling-channel parameters such as pressure, coolant flow velocity, cavity size and the contact angle. The results indicate that the frequency composition of inherent boiling noise in BWRs is not white but lies in the frequency range of 15–45 Hz.

During the last decade a number of design studies on gas cooled fast reactors have been made. They are similar in physics but fundamentally different in design and layout. These differences and their advantages and drawbacks are discussed. So far no single design that combines all the advantages seems possible. Most important is a passive core catcher below the core. Other desirable features include: provision for complete removal of all core components, downward coolant flow, no contact of hot coolant with vessel liner and replaceable pumps and heat exchangers.

KWU pre-Konvoi PWRs (SIEMENS design) are commonly exhibiting high neutron noise levels which can lead to costly operational issues (i.e. activation of SCRAM system, operation of unrated core power, etc.). The frequency region of interest of neutron noise is below 1 Hz, which is the typical frequency range of thermal-hydraulic phenomena. This feature seems to indicate that, coolant flow and temperature oscillations can have a key impact on neutron noise phenomena. Moreover, an increasing neutron noise trend (in term of normalized root mean square) was recently observed in many KWU-PWRs. This increasing trend has been speculated to be correlated with the introduction of a new fuel design type in the KWU-PWRs. This fact indicates that there should be some correlation between neutron noise spectral characteristics and fuel assemblies’ performance. In order to advance in understanding this phenomenon, the transient nodal code SIMULATE-3K (S3K) has been used to simulate mechanical vibrations of fuel assemblies and thermal-hydraulic fluctuations of the core inlet flow and temperature. The simulated neutron detectors responses are analysed with noise analysis techniques and compared to real plant data. This analysis indicates that the cross-feedback between the mechanical and thermal-hydraulic disturbances complicate the identification of the origin of the perturbation source. The simulated results indicate that the neutron noise spectral characteristics can be associated separately to different causes. In this sense, the results of this work seem to indicate that the spectral features of the neutron noise are a consequence of both mechanical perturbations and thermal-hydraulic fluctuations.

The noise that may arise due to boiling of Na in LMFBRs is investigated. The study deals specifically with investigations of the likely frequency range of Na boiling and its dependence on local parameters like coolant flow velocity, cavity size and contact angle.