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: An experiment was performed for the OECD/NEA ROSA-2 Project employing the ROSA/LSTF (rig of safety assessment/large-scale test facility), which simulated a steam generator tube rupture (SGTR) accident induced by main steam line break (MSLB) with operator recovery actions in a pressurized water reactor (PWR). The primary pressure decreased to the pressure level nearly-equal to the intact steam generator (SG) secondary-side pressure even with coolant injection from the high-pressure injection (HPI) system of emergency core cooling system (ECCS) into cold legs. Multi-dimensional coolant behavior appeared such as thermal stratification in both hot and cold legs in intact loop. The RELAP5/MOD3.3 code indicated the insufficient predictions of the primary pressure, the SGTR break flow rate, and the HPI flow rate, and failed to predict the fluid temperatures in the intact loop hot and cold legs. Results obtained from the comparison among three LSTF SGTR-related tests clarified that the thermal stratification occurs in the horizontal legs by different mechanisms.
Abstract: A separate effect test (SET) simulated natural circulation (NC) under high core power condition of a pressurized water reactor (PWR) utilizing the ROSA/LSTF (rig of safety assessment/large-scale test facility). The LSTF test results clarified the relationship between the primary loop mass inventory and the primary loop mass flow rate being dependent on the NC mode at a constant core power of 8% of the volumetric-scaled PWR nominal power. When the core power was 9% or more during reflux condensation, large-amplitude level oscillation in a form of slow fill and dump occurred in steam generator (SG) U-tubes. At 11% core power during reflux condensation, intermittent rise took place in the cladding surface temperature of simulated fuel rods. The RELAP5/MOD3.3 code indicated the insufficient prediction of the SG U-tube liquid level behavior during reflux condensation.
Abstract: ALOHA code was used to calculate the concentration under the CO2 storage burst condition for Maanshan nuclear power plant (NPP) in this study. Five main data are input into ALOHA code including location, building, chemical, atmospheric, and source data. The data from Final Safety Analysis Report (FSAR) and some reports were used in this study. The ALOHA results are compared with the failure criteria of R.G. 1.78 to confirm the habitability of control room. The result of comparison presents that the ALOHA result is below the R.G. 1.78 criteria. This implies that the habitability of control room can be maintained in this case. The sensitivity study for atmospheric parameters was performed in this study. The results show that the wind speed has the larger effect in the concentration calculation.
Abstract: In this research, TRACE code with the interface code-SNAP was used to simulate and analyze the SBO (station blackout) accident which occurred in Maanshan PWR (pressurized water reactor) nuclear power plant (NPP). There are four main steps in this research. First, the SBO accident data of Maanshan NPP were collected. Second, the TRACE/SNAP model of Maanshan NPP was established by using these data. Third, this TRACE/SNAP model was used to perform the simulation and analysis of SBO accident. Finally, the simulation and analysis of SBO with mitigation equipments was performed. The analysis results of TRACE are consistent with the data of Maanshan NPP. The mitigation equipments of Maanshan can maintain the safety of Maanshan in the SBO according to the TRACE predictions.
Abstract: The impact of the storm leads to accidents even in the case of vessels that meet the computed safety criteria for stability. That is why, in order to clarify the causes of the accident and shipwreck, it is necessary to study the dynamics of the ship under the complex sudden impact of external forces. The task is to determine the movement and landing of the ship in the complex and sudden impact of external forces, i.e. when the ship's load changes over a relatively short period of time. For the solution, a technique was used to study the ship's dynamics, which is based on the compilation of a system of differential equations of motion. A coordinate system was adopted for the equation of motion of the hull and the determination of external forces. As a numerical method of integration, the 4th order Runge-Kutta method was chosen. The results of the calculation show that dynamic deviations were lower for high-altitude vessels. The study of the movement of the hull under a difficult situation is performed: receiving of cargo, impact of a flurry of wind and subsequent displacement of the cargo. The risk of overturning and flooding was assessed.
Abstract: In a major nuclear accident, the released fission products (FPs) and the structural materials are likely to influence the transport of iodine in the reactor coolant system (RCS) of a pressurized water reactor (PWR). So far, the thermodynamic data on cesium and silver species used to estimate the magnitude of FP release show some discrepancies, data are scarce and not reliable. For this reason, it is crucial to review the thermodynamic values related to cesium and silver materials. To this end, we have used state-of-the-art quantum chemical methods to compute the formation enthalpies and entropies of AgHMoO₄, CsHMoO₄, and AgCsMoO₄ in the gas phase. Different quantum chemical methods have been investigated (DFT and CCSD(T)) in order to predict the geometrical parameters and the energetics including the correlation energy. The geometries were optimized with TPSSh-5%HF method, followed by a single point calculation of the total electronic energies using the CCSD(T) wave function method. We thus propose with a final uncertainty of about 2 kJmol⁻¹ standard enthalpies of formation of AgHMoO₄, CsHMoO₄, and AgCsMoO₄.
Abstract: Primary containment structure is one of the five safety layers of a nuclear facility which is needed to be designed in such a manner that it can withstand the pressure and excessive radioactivity during accidental situations. It is also necessary to ensure minimization of cost with maximum possible safety in order to make the design economically feasible and attractive. This paper attempts to identify the optimum design conditions for primary containment structure considering both mechanical and radiation safety keeping the economic aspects in mind. This work takes advantage of commercial simulation software to identify the suitable conditions without the requirement of costly experiments. Generated data may be helpful for further studies.
Abstract: In case of abnormal situations, the nuclear power plant (NPP) operators must follow written procedures to check the condition of the plant and to classify the type of emergency. In this paper, we proposed a Real Time Expert System in order to improve operator’s performance in case of transient or accident with reactor shutdown. The expert system’s knowledge is based on the sequence of events (SoE) of known accident and two emergency procedures of the Brazilian Pressurized Water Reactor (PWR) NPP and uses two kinds of knowledge representation: rule and logic trees. The results show that the system was able to classify the response of the automatic protection systems, as well as to evaluate the conditions of the plant, diagnosing the type of occurrence, recovery procedure to be followed, indicating the shutdown root cause, and classifying the emergency level.
Abstract: In this study, we focus on the establishment of the analysis model for Maanshan PWR nuclear power plant (NPP) by using RADTRAD and SNAP codes with the FSAR, manuals, and other data. In order to evaluate the cumulative dose at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) outer boundary, Maanshan NPP RADTRAD/SNAP model was used to perform the analysis of the DBA LOCA case. The analysis results of RADTRAD were similar to FSAR data. These analysis results were lower than the failure criteria of 10 CFR 100.11 (a total radiation dose to the whole body, 250 mSv; a total radiation dose to the thyroid from iodine exposure, 3000 mSv).
Abstract: In this research, the HABIT analysis methodology was established for Maanshan nuclear power plant (NPP). The Final Safety Analysis Report (FSAR), reports, and other data were used in this study. To evaluate the control room habitability under the CO2 storage burst, the HABIT methodology was used to perform this analysis. The HABIT result was below the R.G. 1.78 failure criteria. This indicates that Maanshan NPP habitability can be maintained. Additionally, the sensitivity study of the parameters (wind speed, atmospheric stability classification, air temperature, and control room intake flow rate) was also performed in this research.
Abstract: Spacer grid assembly supporting the nuclear fuel rods
is an important concern in the design of structural components of a
Pressurized Water Reactor (PWR). The spacer grid is composed by
springs and dimples which are formed from a strip sheet by means of
blanking and stamping processes. In this paper, the blanking process
and tooling parameters are evaluated by means of a 2D plane-strain
finite element model in order to evaluate the punch load and quality
of the sheared edges of Inconel 718 strips used for nuclear spacer
grids. A 3D finite element model is also proposed to predict the
tooling loads resulting from the stamping process of a preformed
Inconel 718 strip and to analyse the residual stress effects upon the
spring and dimple design geometries of a nuclear spacer grid.
Abstract: Currently, thorium fuel has been especially noticed
because of its proliferation resistance than long half-life alpha emitter
minor actinides, breeding capability in fast and thermal neutron flux
and mono-isotopic naturally abundant. In recent years, efficiency of
minor actinide burning up in PWRs has been investigated. Hence, a
minor actinide-contained thorium based fuel matrix can confront both
proliferation resistance and nuclear waste depletion aims. In the
present work, minor actinide depletion rate in a CANDU-type nuclear
core modeled using MCNP code has been investigated. The obtained
effects of minor actinide load as mixture of thorium fuel matrix on
the core neutronics has been studied with comparing presence and
non-presence of minor actinide component in the fuel matrix.
Depletion rate of minor actinides in the MA-contained fuel has been
calculated using different power loads. According to the obtained
computational data, minor actinide loading in the modeled core
results in more negative reactivity coefficients. The MA-contained
fuel achieves less radial peaking factor in the modeled core. The
obtained computational results showed 140 kg of 464 kg initial load
of minor actinide has been depleted in during a 6-year burn up in 10
MW power.
Abstract: Fuel rod analysis program transient (FRAPTRAN)
code was used to study the fuel rod performance during a postulated
large break loss of coolant accident (LBLOCA) in Maanshan nuclear
power plant (NPP). Previous transient results from thermal hydraulic
code, TRACE, with the same LBLOCA scenario, were used as input
boundary conditions for FRAPTRAN. The simulation results showed
that the peak cladding temperatures and the fuel centerline
temperatures were all below the 10CFR50.46 LOCA criteria. In
addition, the maximum hoop stress was 18 MPa and the oxide
thickness was 0.003mm for the present simulation cases, which are all
within the safety operation ranges. The present study confirms that this
analysis method, the FRAPTRAN code combined with TRACE, is an
appropriate approach to predict the fuel integrity under LBLOCA with
operational ECCS.
Abstract: This paper will define the system that minimize the risk of the ship accidents because of high or dangerous waves namely early warning system. Since Indonesia is located in a strategic position, many internasional vessels pass by the Indonesian Sea Lanes. Therefore many issues often occur in Indonesian waters, one of the issues is the shipwreck because of dangerous waves. In order to do the preventive action for the vessels that indicated exposed the dangerous waves, Indonesian Maritime Security Coordinating Board or Bakorkamla, has built up and implemented an early warning system through integrated system, called Bakorkamla Integrated Information System (BIIS). By implementing BIIS means that Bakorkamla has already done one of the Five Principles of Sea and Coast Guard Agency, which is safety and security, and Bakorkamla also has already saved the lives of many people on the ship that will have an accident due to high waves.
Abstract: The presented work is motivated by a french law regarding nuclear waste management. In order to avoid the limitation coming with the usage of the existing scenario codes, as COSI, VISION or FAMILY, the Core Library for Advance Scenario Simulation (CLASS) is being develop. CLASS is an open source tool, which allows any user to simulate an electronuclear scenario. The main CLASS asset, is the possibility to include any type of reactor, even a complitely new concept, through the generation of its ACSII evolution database. In the present article, the CLASS working basis will be presented as well as a simple exemple in order to show his potentiel. In the considered exemple, the effect of the transmutation will be assessed on Minor Actinide Inventory produced by PWR reactors.
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 evaluation of unit cell neutronic parameters and
lifetime for some innovant reactors without on sit-refuling will be
held in this work. the behavior of some small and medium reactors
without on site refueling with triso and cermet fuel. For the FBNR
long life except we propose to change the enrichment of the Cermet
MFE to 9%. For the AFPR reactor we can see that the use of the
Cermet MFE can extend the life of this reactor but to maintain the
same life period for AFPR-SC we most use burnup poison to have the
same slope for Kinf (Burnup). PFPWR50 cell behaves almost in
same way using both fuels Cermet and TRISO. So we can conclude
that PFPWR50 reactor, with CERMET Fuel, is kept among the long
cycle reactors and with the new configuration we avoid subcriticality
at the beginning of cycle. The evaluation of unit cell neutronic
parameters reveals a good agreement with the goal of BWR-PB
concept. It is found out that the Triso fuel assembly lifetime can be
extended for a reasonably long period without being refueled,
approximately up to 48GWd/t burnup. Using coated particles fuels
with the Cermet composition can be more extended the fuel assembly
life time, approximately 52 GWd/t.
Abstract: Thermochemcial characteristics of powder fabricated
using oxidation treatment of spent PWR fuel and SIMFUEL were
evaluated for recycling of spent fuel such as DUPIC process.
Especially, the influence of spent fuel burn-ups on the powder
fabrication characteristics was experimentally evaluated, ranging from
27,300 to 65,000 MWd/tU. Densities of powder manufactured from an
oxidation, OREOX and the milling processes at the same process
conditions were compared as a function of the fuel burn-ups
respectively. Also, based on chemical analysis results, homogeneity of
fissile elements in oxidized powder was confirmed.