Primary Coolant Pumps Research Articles

Design of control strategy for reactor power control system of LBE cooled Actinide Burner Reactor (ABR) using system analysis code NUSOL-LMR

The power control strategy for the Lead–Bismuth Eutectic (LBE) cooled Actinide Burner Reactor (ABR) has been designed to assess the closed-loop performance of newly developed control blocks that were not modeled in the NUSOL-LMR code before this research effort. After establishing a sustainable model of ABR in NUSOL-LMR, validation of the model in light of the already-published research is performed. The said closed-loop control strategy (consisting of the power and temperature control blocks) is evaluated under different transient scenarios (reactivity insertion, primary coolant pump fault, and change in demand power levels). The power and temperature control blocks concurrently control reactor power during any transient occurrence. The coupling effect of the primary coolant temperature deviation on the reactor power has been minimized by the proportional and integral (PI) controller in the temperature control block. The simulation results demonstrate that the closed-loop performance of the ABR under the controller action during transient events is quite satisfactory. The closed-loop power regulation performance of the conventional sliding mode controller (SMC) with a PI sliding surface (PISM) in the power control block of NUSOL-LMR is better than that of a simple proportional, integral and derivative (PID) controller. The asymptotic stability of the closed-loop system with the PISM controller has been ensured using the Lyapunov stability theory.

Safety Analysis in AP-1000 Nuclear Power Plant Against Primary Loop Coolant Pump and Fuel Failures Authors

Safety Analysis in AP-1000 Nuclear Power Plant Against Primary Loop Coolant Pump and Fuel Failures Authors

Validation of the thermal-hydraulic thunder code and application for PWR core in transient conditions

The THUNDER computer code allows performing thermal-hydraulic core analysis of pressurized water reactors aiming at the design and safety analysis of power and naval reactors in steady state and transient conditions. It uses the subchannel and control volume methods to perform a tridimensional analysis of the nuclear reactor core. The conservation equations are solved simultaneously for each control volume based on geometric and operation conditions. The Integral Momentum Method is used for solving the conservation equations decoupling the momentum equation from the mass and energy conservation equations. The drift-flux model is adopted to treat the two-phase flow condition. The analysis is performed in two-stages: first, a macroscopic analysis is conducted with subchannel equivalent to a fuel assembly (bundle), and second, with a subchannel equivalent to an individual coolant channel. Only the hottest fuel assembly is analyzed with detail. To validate the THUNDER code we used the experimental data from the NUPEC PWR benchmark (subchannel and bundle tests), in which we verify the accuracy of the coolant temperature at steady-state in 4 different void fraction conditions and the accuracy of the void fraction at transients initiated by four different conditions: core power increase, reduction of inlet mass flow rate, increase of inlet coolant temperature and depressurization of the inlet coolant. At steady-state conditions, the mean relative discrepancy for the coolant temperature was 0.09. At transient conditions, the void fraction mean absolute discrepancy for power increase was 0.045, for mass flow reduction was 0.079, for inlet coolant depressurization was 0.032 and for temperature increase was 0.092. The code was applied for the Brazilian 1350 MW nuclear power plant ANGRA 2 for evaluating the thermal-hydraulic safety parameters at steady-state, as well as the operational transients of loss of one or several primary coolant pumps. The THUNDER code results showed that the safety limits were not exceeded in concordance to the Final Safety Analysis Report of the plant. These results indicate that the THUNDER code is sufficiently accurate to perform thermal-hydraulic analyses for PWR power plants at steady-state and transient conditions.

STEADY-STATE ANALYSIS ON COOLANT SYSTEM OPERATING OF THE TRIGA2000 USING FUEL PLATE TYPE

The TRIGA2000 is one of the nuclear research reactors in Indonesia. TRIGA reactor fuel element is rod type, studies of the TRIGA reactor using plate type fuel have been carried out. One primary coolant pump is required to overcome the safety factor on the core which uses plate-type fuel. Besides that, changes in the reactor core configuration have an effect on the ability to transfer heat from the core to the coolant. Therefore, it is necessary to analyze the operating parameters of the cooling system including temperature, pressure and flow rate. This operating data is important, which is required by the operator to estimate the reactor coolant condition. The purpose of this paper is to analyze the cooling system parameters under normal steady-state operation where the primary coolant capacity is 50 kg/s and the reactor power is determined to be 1 MW. Calculation of cooling system at normal condition is carried out using the ChemCAD.6.1.4 computer program. This is also required in providing the reactor's Safety Analysis Report (SAR) document. The results show that the TRIGA reactor which uses plate-type fuel elements can be operated safely. However, in case of coolant temperature to the reactor core exceeding the design data (35 o C), it is necessary to be considered against the operational reliability of the cooling system

Fast Defect Detection on Primary Pump Pipe for RSG-GAS Reactor Using Acoustics Emission Techniques

RSG-GAS reactor is one of the research reactors owned by National Nuclear Energy Agency of Indonesia (BATAN) which has been operated since 1987 and utilized its neutrons for radioisotope production and other research activities, and heat caused by nuclear fission reactions is discharged into the environment. To remove the heat is required coolant pumps taking heat from the reactor core to the environment. The RSG-GAS coolant pump consists of a primary coolant and secondary coolant with a heat transfer capacity of 33,000 kW and 5,500 kW respectively. The primary coolant pump takes heat directly from the core and the heat is then transported to the secondary coolant pump system as it passes through the heat exchanger. It then discharges the environment by the cooling system secondary on the cooling tower by blowing using a blower that works counter flow to the direction of the coolant flow. The drainage from the reactor core into the environment requires pipeline from the component system to other components at the pump. Since RSG-GAS reactor has reached 30 years, the RSG-GAS reactor coolant system has aging experienced and decrease of its performance. The occurrence of sound and vibration in the cooling primary pipe indicated a defect location or cavitation due to hard impact of water flow and exposure of gamma radiation on the pipe wall continuously during reactor operation. To ensure the occurrence of defect location, it is necessary to conduct research and testing on the material structure of the cooling pump pipe. Testing and research on the structure of the cooling pipe material can not be done by disassembling the pipes because it is installed firmly and is at risk of high exposure gamma radiation. It is, therefore, nondestructive test method (NDT), one of them using acoustic emission method, should be applied. The experiments to detect the location defect on pipe of primary cooling pump of RSG-GAS reactor have been conducted. The results on this research showed primary cooling pump pipe degradation and low cavitation existence.

Code validation of a scaled-down DRACS model in RELAP5/SCDAPSIM/MOD 4.0

Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system proposed for Fluoride-salt-cooled High-temperature Reactors (FHRs). Benchmark study of decay heat removal capability of DRACS by natural circulation is critical for FHR safety analysis. Two scaled-down test facilities, a low-temperature DRACS test facility (LTDF) and a high-temperature DRACS test facility (HTDF), were designed following a scaling analysis and constructed to investigate thermal performance and heat removal capability of the DRACS. The LTDF uses water as a surrogate coolant for both the primary and DRACS loops. Two transient scenarios were carried out in the LTDF, namely, DRACS startup and primary coolant pump trip. In both of the scenarios, the experimental results demonstrate sufficient heat removal capabilities of the DRACS with natural circulations established in the system. These two scenarios were simulated using RELAP5/SCDASIM/MOD4.0 code and the code simulation results show good agreement with the LTDF experimental data. For the HTDF, molten salts FLiNaK and KF-ZrF4 are used respectively as the primary coolant and DRACS coolant. Thermodynamic and transport properties of FLiNaK and KF-ZrF4 were implemented into RELAP5/SCDAPSIM/MOD 4.0. Simulations of transient scenarios in the HTDF were also performed and the code results will be compared with the HTDF experimental data once available.

Simulation of Modified TRIGA-2000 with Plate-Type Fuel under LOFA Using EUREKA2/RR-Code

The TRIGA-2000 research reactor in Bandung, Indonesia, has operated for over 50 years. Recently, the problem of fuel availability arises, since its fuel is no longer produced. A modification of reactor core with new plate-type fuel has been suggested . The study of the neutronic assessment of plate-type fuel elements reactor core had been done. The next assessment that needed to be done was thermal-hydraulic analysis. The purpose of this study is to simulate the thermal-hydraulic characteristics of major parameters, such as reactor power, fuel cladding temperature, and departure from nucleate boiling ratio (DNBR) due to LOFA transient, using EUREKA2/RR code. During steady-state condition, downward flow forced convection mode for core coolant system is operated. The upward flow occurs when the natural circulation mode takes place. Hottest core channels temperature during LOFA conditions was considered. The reactor core was modeled as three channels, i . e ., the hottest channel, the average channel, and the channel for control assemblies, respectively. The s imulation was based on the steady-state condition of 2 M W t reactor power, cooling mass flow rate of 63.5 kg/s , and inlet coolant temperature to the core of 3 5.5 °C. The result shows that the hottest fuel cladding temperature does not cause a nucleate boiling. During LOFA, the residual heat was removed by natural circulation flow that occurred slowly. In order t o have larger inertia force , provision of the flywheel in the shaft of primary coolant pump is suggested.

Acoustic measurement infrastructure to enable the enhancement of nuclear reactor efficiency and safety

Nuclear research reactors are used to test efficiency and safety of new technology, such as material or fuel samples under prototypic commercial conditions, operation of reactor prototypes, safety studies, assessment of neutron parameters, etc. Many experiments performed in research reactors require in-situ measurements to monitor the progress and performance of conducted tests. The core of any nuclear reactor presents a particularly harsh environment for sensors and instrumentations. The reactor core also imposes challenging constraints on signal transmission from inside the reactor core to outside of the reactor vessel. In this paper, an acoustic measurement infrastructure (AMI) installed at the Advanced Test Reactor (ATR), located at Idaho National Laboratory, is presented. The AMI consists of ATR in-pile structural components, coolant, acoustic receivers, primary coolant pumps, a data-acquisition system, and signal processing algorithms. Intrinsic and cyclic acoustic signals generated by the operation of the primary coolant pumps are collected using acoustic receivers and processed. The characteristics of the intrinsic signal can indicate the process state of the ATR during operation (i.e., real-time measurement). The innovation of AMI can be extended to collect information on other phenomena, such as fuel motion, individual fuel rod vibration, loose parts, thermal expansion, and flow blockage, that occur inside an operating nuclear reactor.

The research on AP1000 nuclear main pumps’ complete characteristics and the normalization method

The paper summarizes the complete characteristics of nuclear main pumps based on experimental results and makes a detailed study, and then draws a series of important conclusions: with regard to the overall flow area, the runaway operating and 0-revolving-speed operating of nuclear main pumps both have quadratic characteristics; with regard to the infinite flow, the braking operation and the 0-revolving-speed operation show consistent external characteristics. To remedy the shortcomings of the traditional complete-characteristic expression with regards to only describing limited flow sections at specific revolving speeds, the paper proposes a normalization method. As an important boundary condition of the security computing of unstable transient process of the primary reactor coolant pump and the nuclear island primary circuit and secondary circuit, the precision of complete-characteristic data and curve impacts the precision of security computing. A normalization curve obtained by applying the normalization method to process complete-characteristic data could correctly, completely and precisely express the complete characteristics of the primary reactor coolant pump under any rotational speed and full flow, and is capable of giving proper estimations on external characteristics of the flow outside the test range and even of the infinite flow. These advantages are of great significance for the improvement of security computing of transient processes of the primary reactor coolant pump and the circuit system.

Thermal Aging Phenomena in Cast Duplex Stainless Steels

Cast stainless steels (CASSs) have been extensively used for the large components of light water reactor (LWR) power plants such as primary coolant piping and pump casing. The thermal embrittlement of CASS components is one of the most serious concerns related to the extended-term operation of nuclear power plants. Many past researches have concluded that the formation of Cr-rich α′-phase by Spinodal decomposition of δ-ferrite phase is the primary mechanism for the thermal embrittlement. Cracking mechanism in the thermally-embrittled duplex stainless steels consists of the formation of cleavage at ferrite and its propagation via separation of ferrite–austenite interphase. This article intends to provide an introductory overview on the thermal aging phenomena in LWR-relevant conditions. Firstly, the thermal aging effect on toughness is discussed in terms of the cause of embrittlement and influential parameters. An approximate analysis of thermal reaction using Arrhenius equation was carried out to scope the aging temperatures for the accelerated aging experiments to simulate the 60 and 80 years of services. Further, an equilibrium precipitation calculation was performed for model CASS alloys using the CALPHAD program, and the results are used to describe the precipitation behaviors in duplex stainless steels. These results are also to be used to guide an on-going research aiming to provide knowledge-based conclusive prediction for the integrity of the CASS components of LWR power plants during the service life extended up to and beyond 60 years.

Experimental and analytical investigations of primary coolant pump coastdown phenomena for the Jordan Research and Training Reactor

Many low-power research reactors including the Jordan Research and Training Reactor (JRTR) are designed to have a downward core flow during a normal operation mode for many convenient operating features. This design feature requires maintaining the downward core flow for a short period of time right after a loss of off-site power (LOOP) accident to guarantee nuclear fuel integrity. In the JRTR, a big flywheel is installed on a primary cooling system (PCS) pump shaft to passively provide the inertial downward core flow at an initial stage of the LOOP accident. The inertial pumping capability during the coastdown period is experimentally investigated to confirm whether the coastdown half time requirement given by safety analyses is being satisfied. The validity and consistency of the experimental dataset are evaluated using a simulation software package, modular modeling system (MMS). In the MMS simulation model, all of the design data that affect the pump coastdown behavior are reflected. The experimental dataset is well predicted by the MMS model, and is confirmed to be valid and consistent. The designed coastdown half time is confirmed to be well above the value required by safety analysis results.wwwyoon@gmail.com

Analyses for Primary Coolant Pump Coastdown Phenomena for Jordan Research and Training Reactor

Analyses for Primary Coolant Pump Coastdown Phenomena for Jordan Research and Training Reactor

Analysis of Reactivity Insertion Accidents for the NIST Research Reactor before and after Fuel Conversion

Reactivity insertion accidents have been analyzed for the 20-MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains highly enriched uranium fuel, and for a proposed equilibrium core with low-enriched uranium fuel. The time-dependent analysis of the primary system is performed with a RELAP5 model that includes the reactor vessel, primary coolant pump, heat exchanger, fuel element geometry, and flow channels for both the 6 inner and 24 outer fuel elements. Postprocessing of the simulation results has been conducted to evaluate minimum critical heat flux (CHF) ratio and minimum onset of flow instability (OFI) ratio using the Sudo-Kaminaga correlations and Saha-Zuber criteria, respectively.Evaluations are carried out for the control rod withdrawal start-up accident and the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that no damage to the fuel will occur and there is adequate margin to CHF and OFI because of sufficient coolant flow through the fuel channels and the negative reactivity insertion due to scram.

Thermal hydraulic analysis of the AHWR—The Indian thorium fuelled innovative nuclear reactor

Analysis has been carried out for simulating loss-of-coolant accident (LOCA) at inlet header in a natural circulation type reactor developed as the advanced heavy water reactor (AHWR).The paper will cover a case of LOCA due to 200% break at inlet header which is double ended rupture. The maximum clad surface temperature has been predicted in different cases by using the thermal hydraulic safety code RELAP5/Mod4.0. The proposed reactor is a 920MWth vertical pressure tube type, boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is heat removal through natural circulation of primary coolant (at all allowed power levels) with no primary coolant pumps. This reactor is equipped with emergency core cooling system (ECCS) and isolation condensers (ICs) to remove decay heat during LOCA. This ECCS provides cooling to fuel in passive mode during first fifteen minutes of LOCA and it is achieved by high pressure injection from advanced accumulator. Cooling is continued for Later for three days by the gravity driven water pool (GDWP). This paper investigates the impact of high pressure injection in this cooling process.

Simulations of RUTA-70 reactor with CERMET fuel using DYN3D/ATHLET and DYN3D/RELAP5 coupled codes

Abstract RUTA-70 model for simulations with the internally coupled codes DYN3D/ATHLET and DYN3D/RELAP5 was developed. A 3-D power distribution in the core is calculated by DYN3D with thermal-hydraulic feedback from the system codes. A steady-state corresponding to the full reactor power and an accident scenario initiated by failure of all primary coolant pumps were simulated with the DYN3D/ATHLET and DYN3D/RELAP5 coupled code systems to verify these codes. The compared coupled codes give close predictions for the initial and final states of the simulated accident but not for the transition between them. The observed deviations are explained by differences in the subcooled boiling models of the employed versions of ATHLET and RELAP5. Nevertheless, both simulations confirm a high level of the reactor inherent safety. The allowed safety margins were not reached.

Numerical Simulation of Thermal Stratification in Cold Legs by Using OpenFOAM

*,† During a small-break loss-of-coolant accident in pressurized water reactors (PWRs), emergency core cooling system (ECCS) is actuated and cold water is injected into cold legs. Insufficient mixing of injected cold water and hot primary coolant results in thermal stratification, which is a matter of concern for evaluation of pressurized thermal shock (PTS) in view of aging and life extension of nuclear power plants. In this study, an open source CFD software, OpenFOAM, is used to simulate mixing and thermal stratification in the cold leg of ROSA/LSTF, which is the largest thermal-hydraulic integral test facility simulating PWR. One of the cold-leg is numerically simulated from the outlet of primary coolant pump to the inlet of downcomer. ECCS water is injected from injection nozzle connected at the top of the cold leg into the steady-state natural circulation flow under high-pressure and high-temperature conditions. The temperature distribution in the cold leg is compared with experimental and FLUENT’s results. Effects of turbulent flow models and secondary flow due to the elbow section of the cold leg are discussed for the case with the single-phase natural circulation. Injection into a two-phase stratified flow is also simulated and predictive and numerical capabilities of OpenFOAM are discussed.

Simulation of load following mode of operation for a natural circulation pressure tube type BWR

Abstract A new nuclear power reactor under design study is a vertical pressure tube type boiling light water cooled and heavy water moderated. One of the passive design features of this reactor is the heat removal through natural circulation of primary coolant at all power level with no primary coolant pumps. Nuclear plants are mainly base load units, but the proposed plant with various advance features has to operate in load following mode i.e. Reactor follows Turbine (in a limited range). In this mode, any alteration in turbine load results in the steam pressure change. The steam pressure error is fed to the Reactor Regulating System (RRS), which changes the reactor power to control the system pressure. To study this mode of plant operation, a plant simulation model with the feedbacks from various controllers has been developed using the RELAP5 code. This integrated plant model has been used for simulating the load-varying scenario for a change in plant load. All the process dynamics, modeling, design verification and performance issues are discussed in this paper.

Possible design improvements and a high power density fuel design for integral type small modular pressurized water reactors

The study evaluates potential weaknesses and possible improvements for integral type small modular pressurized water reactor designs. By taking International Reactor Innovative and Secure (IRIS) as the reference design and keeping the power output as the same, a new fuel and reactor design were proposed. The proposed design relocates the primary coolant pumps and the pressurizer outside the reactor pressure vessel (RPV). Three recirculation lines and jet pumps/centrifugal pumps are introduced to provide the coolant circulation similar to Boiling Water Reactor designs. The pressurizer component is expected to be similar to the AP600 design. It is located at one of the recirculation lines. The new fuel assembly adopts 264 solid cylindrical fuel pins with 10 mm diameter and 2.3 m height, arranged at a hexagonal tight lattice configuration. Large water rods are introduced to preserve the moderating power and to accommodate finger type control rods. The resulting fuel can operate with 104.5 kW/l power density while having substantially higher margin for boiling crisis compared to typical large PWRs. Full core neutronic analysis shows that 24-month cycle length and 50 MWd/kg burnup is achievable with a two-batch refueling scheme. Furthermore, the fuel behavior study shows that the new fuel with M5 type Zircaloy cladding show fairly acceptable steady state performance. A preliminary Loss of Coolant analysis shows that the new design could be advantageous over IRIS due to its low ratio of the water inventory below the top of the active fuel to total RPV water inventory. The proposed reactor pressure vessel height and the containment volume are 30% lower than the reference IRIS design.

Thermal hydraulic analysis for grid supported pressurized water reactor cores

This paper presents the methodology and results for thermal hydraulic analysis of grid supported pressurized water reactor cores using U(45% wt)–ZrH 1.6 hydride fuel in square arrays. The same methodology is applied to the design of UO 2 oxide fueled cores to provide a fair comparison of the achievable power between the two fuel types. Steady-state and transient design limits are considered. Steady-state limits include: fuel bundle pressure drop, departure from nucleate boiling ratio, fuel temperature (average for UO 2 and centerline/peak for U–ZrH 1.6), and fuel rod vibrations and wear. Transient limits are derived from consideration of the loss of flow and loss of coolant accidents, and an overpower transient. In general, the thermal hydraulic performance of U–ZrH 1.6 and UO 2 fuels is very similar. Slight power differences exist between the two fuel types for designs limited by rod vibrations and wear, because these limits are fuel dependent. Large power increases are achievable for both fuels when compared to the reference core power output of 3800 MW th. In general, these higher power designs have smaller rod diameters and larger pitch-to-diameter ratios than the reference core geometry. If the pressure drop across new core designs is limited to the pressure drop across the reference core, power increases of ∼400 MW th may be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure drop equal to twice the reference core pressure drop, power increases of ∼1000 MW th may be feasible.

Thermal hydraulic effect of fuel plate surface roughness

This study presents surface roughness measurements characteristic of the pre-film layer applied to a typical Advanced Test Reactor (ATR) fuel plate. This data is used to estimate the friction factor for thermal hydraulic flow calculations of a Gas Test Loop (GTL) system proposed for incorporation into ATR to provide a fast neutron flux environment for the testing of nuclear fuels and materials. To attain the required neutron flux, the design includes booster fuel plates clad with the same aluminum alloy as the ATR driver fuel and cooled with water supplied by the ATR primary coolant pumps. The objectives of this study are to (1) determine the surface roughness of the protective boehmite layer applied to the ATR driver fuel prior to reactor operations in order to specify the machining tolerances for the surface finish on simulated booster fuel plates in a GTL hydraulic flow test model and (2) assess the consequent thermal hydraulic impact due to surface roughness on the coolability of the booster fuel with a similar pre-film layer applied. While the maximum roughness of this coating is specified to be 1.6 μm (63 μin.), no precise data on the actual roughness were available. A representative sample coupon autoclaved with the ATR driver fuel to produce the pre-film coating was analyzed using optical profilometry. Measurements yielded a mean surface roughness of 0.53 μm (21 μin.). Results from a sensitivity study show that a ±15% deviation from the mean measured surface finish would have a minimal effect on coolant temperature, coolant flow rate, and fuel temperature. However, frictional losses from roughnesses greater than 1.5 μm (∼60 μin.) produce a marked decrease in flow rate, causing fuel and coolant temperatures to rise sharply.