Abstract: This work presents the particle swarm optimization trained neural network (PSO-NN) supervisory proportional integral derivative (PID) control method to monitor the pressurized water reactor (PWR) core power for safe operation. The proposed control approach is implemented on the transfer function of the PWR core, which is computed from the state-space model. The PWR core state-space model is designed from the neutronics, thermal-hydraulics, and reactivity models using perturbation around the equilibrium value. The proposed control approach computes the control rod speed to maneuver the core power to track the reference in a closed-loop scheme. The particle swarm optimization (PSO) algorithm is used to train the neural network (NN) and to tune the PID simultaneously. The controller performance is examined using integral absolute error, integral time absolute error, integral square error, and integral time square error functions, and the stability of the system is analyzed by using the Bode diagram. The simulation results indicated that the controller shows satisfactory performance to control and track the load power effectively and smoothly as compared to the PSO-PID control technique. This study will give benefit to design a supervisory controller for nuclear engineering research fields for control application.
Abstract: The aim of this paper is to perform, by mean of the finite volume method, a numerical solution of the transient natural convection in a narrow rectangular channel between two vertical parallel Material Testing Reactor (MTR)-type fuel plates, imposed under a heat flux with a cosine shape to determine the margin of the nuclear core power at which the natural convection cooling mode can ensure a safe core cooling, where the cladding temperature should not reach a specific safety limits (90 °C). For this purpose, a computer program is developed to determine the principal parameters related to the nuclear core safety, such as the temperature distribution in the fuel plate and in the coolant (light water) as a function of the reactor core power. Throughout the obtained results, we noticed that the core power should not reach 400 kW, to ensure a safe passive residual heat removing from the nuclear core by the upward natural convection cooling mode.
Abstract: The aim of this paper is to perform a thermal-hydraulic analysis of the IAEA 10 MW benchmark reactor solving analytically and numerically, by mean of the finite volume method, respectively the steady state and transient forced convection in rectangular narrow channel between two parallel MTR-type fuel plates, imposed under a cosine shape heat flux. A comparison between both solutions is presented to determine the minimal coolant velocity which can ensure a safe reactor core cooling, where the cladding temperature should not reach a specific safety limit 90 °C. For this purpose, a computer program is developed to determine the principal parameter related to the nuclear core safety, such as the temperature distribution in the fuel plate and in the coolant (light water) as a function of the inlet coolant velocity. Finally, a good agreement is noticed between the both analytical and numerical solutions, where the obtained results are displayed graphically.
Abstract: Heat pipe is considered to be applied as a passive system to remove residual heat that generated from reactor core when incident occur or from spent fuel storage pool. The objectives are to characterized the heat transfer phenomena, performance of heat pipe, and as a model for large heat pipe will be applied as passive cooling system on nuclear spent fuel pool storage. In this experimental wickless heat pipe or two-phase closed thermosyphon (TPCT) is used. Variation of heat flux are 611.24 Watt/m2 - 3291.29 Watt/m2. Variation of filling ratio are 45 - 70%. Variation of initial pressure are -62 to -74 cm Hg. Demineralized water is used as working fluid in the TPCT. The results showed that increasing of heat load leads to an increase of evaporation of the working fluid. The optimum filling ratio obtained for 60% of TPCT evaporator volume, and initial pressure variation gave different TPCT wall temperature characteristic. TPCT showed best performance with 60% filling ratio and can be consider to be applied as passive residual heat removal system or passive cooling system on spent fuel storage pool.
Abstract: This paper presents the modeling approach in SBO
sequence for VVER 1000 reactors and describes the reactor core
behavior at late in-vessel phase in case of late reflooding by HPIS
and gives preliminary results for the ASTECv2 validation. The work
is focused on investigation of plant behavior during total loss of
power and the operator actions. The main goal of these analyses is to
assess the phenomena arising during the Station blackout (SBO)
followed by primary side high pressure injection system (HPIS)
reflooding of already damaged reactor core at very late “in-vessel”
phase. The purpose of the analyses is to define how the later HPIS
switching on can delay the time of vessel failure or possibly avoid
vessel failure. The times for HPP injection were chosen based on
previously performed investigations.
Abstract: An appropriate model to predict the size of the droplets
resulting from the break-up with the structures will help in a better
understanding and modeling of the two-phase flow calculations in the
simulation of a reactor core loss-of-coolant accident (LOCA). A
droplet behavior impacting on a hot surface above the Leidenfrost
temperature was investigated. Droplets of known size and velocity
were impacted to an inclined plate of hot temperature, and the
behavior of the droplets was observed by a high-speed camera. It was
found that for droplets of Weber number higher than a certain value,
the higher the Weber number of the droplet the smaller the secondary
droplets. The COBRA-TF model over-predicted the measured
secondary droplet sizes obtained by the present experiment. A simple
model for the secondary droplet size was proposed using the mass
conservation equation. The maximum spreading diameter of the
droplets was also compared to previous correlations and a fairly good
agreement was found. A better prediction of the heat transfer in the
case of LOCA can be obtained with the presented model.
Abstract: Detailed thermal hydraulic investigations are very
essential for safe and reliable functioning of liquid metal cooled fast
breeder reactors. These investigations are further more important for
components with complex profile, since there is no direct correlation
available in literature to evaluate the hydraulic characteristics of such
components directly. In those cases available correlations for similar
profile or geometries may lead to significant uncertainty in the
outcome. Hence experimental approach can be adopted to evaluate
these hydraulic characteristics more precisely for better prediction in
reactor core components.
Prototype Fast Breeder Reactor (PFBR), a sodium cooled pool
type reactor is under advanced stage of construction at Kalpakkam,
India. Several components of this reactor core require hydraulic
investigation before its usage in the reactor. These hydraulic
investigations on full scale models, carried out by experimental
approaches using water as simulant fluid are discussed in the paper.
Abstract: In this research, the TRACE/PARCS model of
Lungmen ABWR has been developed for verification of ultimate
response guideline (URG) efficiency. This ultimate measure was
named as DIVing plan, abbreviated from system depressurization,
water injection and containment venting. The simulation initial
condition is 100% rated power/100% rated core flow. This research
focuses on the estimation of the time when the fuel might be damaged
with no water injection by using TRACE/PARCS first. Then, the
effect of the reactor core isolation system (RCIC), control
depressurization and ac-independent water addition system (ACIWA),
which can provide the injection with 950 gpm are also estimated for
the station blackout (SBO) transient.
Abstract: This article summarizes ways to verify neutron
fluence for neutron transmutation doping of silicon with phosphorus
on the LVR-15 reactor. Neutron fluence is determined using
activation detectors placed along the crystal in a strip or encapsulated
in a rod holder. Holders are placed at the centre of a water-filled
capsule or in an aluminum or silicon ingot that simulates a real single
crystal. If the diameter of the crystal is significantly less than the
capsule diameter and water from the primary circuit enters the free
space in the capsule, neutron interaction in the water changes neutron
fluence, affecting axial irradiation homogeneity. The effect of
moving the capsule vertically in the channel relative to maximum
neutron fluence in the reactor core was also measured. Even a small
shift of the capsule-s centre causes great irradiation inhomogeneity.
This effect was measured using activation detectors, and was also
confirmed by MCNP calculation.
Abstract: Most HWRs currently use natural uranium fuel. Using enriched uranium fuel results in a significant improvement in fuel cycle costs and uranium utilization. On the other hand, reactivity changes of HWRs over the full range of operating conditions from cold shutdown to full power are small. This reduces the required reactivity worth of control devices and minimizes local flux distribution perturbations, minimizing potential problems due to transient local overheating of fuel. Analyzing heavy water effectiveness on neutronic parameters such as enrichment requirements, peaking factor and reactivity is important and should pay attention as primary concepts of a HWR core designing. Two nuclear nuclear reactors of CANDU-type and hexagonal-type reactor cores of 33 fuel assemblies and 19 assemblies in 1.04 P/D have been respectively simulated using MCNP-4C code. Using heavy water and light water as moderator have been compared for achieving less reactivity insertion and enrichment requirements. Two fuel matrixes of (232Th/235U)O2 and (238/235U)O2 have been compared to achieve more economical and safe design. Heavy water not only decreased enrichment needs, but it concluded in negative reactivity insertions during moderator density variations. Thorium oxide fuel assemblies of 2.3% enrichment loaded into the core of heavy water moderator resulted in 0.751 fission to absorption ratio and peaking factor of 1.7 using. Heavy water not only provides negative reactivity insertion during temperature raises which changes moderator density but concluded in 2 to 10 kg reduction of enrichment requirements, depend on geometry type.
Abstract: The paper investigates parallel channel instabilities of
natural circulation boiling water reactor. A thermal-hydraulic model
is developed to simulate two-phase flow behavior in the natural circulation boiling water reactor (NCBWR) with the incorporation of
ex-core components and recirculation loop such as steam separator, down-comer, lower-horizontal section and upper-horizontal section
and then, numerical analysis is carried out for parallel channel
instabilities of the reactor undergoing both in-phase and out-of-phase
modes of oscillations. To analyze the relative effect on stability of the reactor due to inclusion of various ex-core components and
recirculation loop, marginal stable point is obtained at a particular inlet enthalpy of the reactor core without the inclusion of ex-core
components and recirculation loop and then with the inclusion of the
same. Numerical simulations are also conducted to determine the
relative dominance between two modes of oscillations i.e. in-phase and out-of-phase. Simulations are also carried out when the channels
are subjected to asymmetric power distribution keeping the inlet enthalpy same.
Abstract: For a quick and accurate calculation of spatial neutron
distribution in nuclear power reactors 3D nodal codes are usually
used aiming at solving the neutron diffusion equation for a given
reactor core geometry and material composition. These codes use a
second order polynomial to represent the transverse leakage term. In
this work, a nodal method based on the well known nodal expansion
method (NEM), developed at COPPE, making use of this polynomial
expansion was modified to treat the transverse leakage term for the
external surfaces of peripheral reflector nodes.
The proposed method was implemented into a computational
system which, besides solving the diffusion equation, also solves the
burnup equations governing the gradual changes in material
compositions of the core due to fuel depletion. Results confirm the
effectiveness of this modified treatment of peripheral nodes for
practical purposes in PWR reactors.
Abstract: The presented paper is related to the design methods and neutronic characterization of the reactivity control system in the large power unit of Generation IV Gas cooled Fast Reactor – GFR2400. The reactor core is based on carbide pin fuel type with the application of refractory metallic liners used to enhance the fission product retention of the SiCcladding. The heterogeneous design optimization of control rod is presented and the results of rods worth and their interferences in a core are evaluated. In addition, the idea of reflector removal as an additive reactivity management option is investigated and briefly described.