Chemical And Volume Control System Research Articles

Multiscale analysis of the boron dilution sequence in the NuScale reactor using TRACE and SUBCHANFLOW

In the ‘High-Performance Advanced Methods and Experimental Investigations for the Safety Evaluation of Generic Small Modular Reactors’ McSAFER H2020 European project the main aim pursued is to advance the research in safety analysis methods for water cooled Small Modular Reactors (SMRs). Hence, several light-water-cooled SMR concepts were selected along with the corresponding accident scenarios to be modelled to evaluate the potential application of advanced modeling tools based on the coupling of different codes and their benefits and drawbacks. In this study, the simulation of the postulated boron dilution scenario caused by the malfunction of the Chemical and Volume Control System (CVCS) in the NuScale reactor is performed using different modeling tools and approaches. More in particular, a TRACE-1D, a TRACE-3D and a TRACE-3D/SUBCHANFLOW (SCF) models have been developed to perform this analysis. Two cases have been carried out in this work: the first case is based on the calculation of the boron concentration evolution within the Reactor Coolant System (RCS) of NuScale to compute the time to loss of shutdown margin (SDM), and secondly, a best estimate calculation is performed to evaluate the plant response considering the reactivity feedbacks and the performance of the safeguards and safety systems of NuScale according to the public literature. These investigations have demonstrated that the applied modeling approaches and tools can capture the physics of the problem providing less conservative results than those obtained using the analytical methods presented in the DCA report of NuScale. In addition, it has been demonstrated that the NuScale design is robust against the consequences of the boron dilution transient and that no relevant asymmetrical effects appears due to the singular arrangement of the Helically Coiled Steam Generators (HCSGs) surrounding the riser.

Analysis of depressurization ability for IRIS containment during SBLOCAs using RELAP5 and CONTEMPT

As an integral and modular pressurized water reactor (PWR), the international reactor innovative and secure (IRIS) eliminates the threat of large break LOCAs (LBLOCAs) by integrating all major reactor coolant system components into the reactor vessel and provides compact steel containment with a suppression pool. Unlike large reactor containment, IRIS containment has modest volume and compact internal structure. The analysis of traditional dry containment cannot be applied to the IRIS reactor. Despite the disappearance of LBLOCAs (exceeding 4 in), small steel containment still faces the threat of high–frequency small break LOCAs (SBLOCAs) (e.g., the largest possible 4 in chemical and volume control system (CVCS) line break and the lowest elevation 2 in direct vessel injection (DVI) line break). To identify the impact of these characteristics on the integrity of IRIS reactor containment, a safety assessment framework that combines deterministic and probabilistic methods is proposed for the pressure response of containment under SBLOCAs of IRIS reactor. First in the modeling phase, primary and secondary loop models have been built considering the design of IRIS, and accident consequences under typical SBLOCAs were evaluated using RELAP5/MODE4. A simplified model of containment was presented herein based on the primary structure of IRIS containment. Second in the input parameters identification and uncertainty quantification stage, the responses of containment were simulated using CONTEMPT–LT. Thus, the main factors affecting the depressurization capacity of the containment were identified, and the reasonable probability density distributions were determined. Third, in the uncertainty propagation stage, Monte Carlo method was used to map the uncertainty of input parameters to the output variable. Fourth in the sensitivity analysis and uncertainty analysis phase, the contribution of each input parameter was measured using sensitivity analysis, and their interactions were investigated. The range of deviations of the containment pressure response was measured using uncertainty analysis, and finally, the calculated results were combined with the containment fragility curve to evaluate containment failure probability.

Transient Analysis of CVCS Malfunction in Large Passive PWR

The present work focuses on a simulation of increase of coolant inventory due to malfunction of chemical and volume control system (CVCS) in advance passive pressure water reactor 1000 (AP1000) operating at a 102% nominal power before transient start-up. The objective is to verify if this transient may induce any over pressure in the reactor coolant pressure boundary due to inadequate residual heat removal. RELAP5 system code was used to simulate the transient and the simulation result was compared with the simulation result by LOFTRAN code. The RELAP5 simulation result is well comparable with the simulation result by LOFTRAN and this transient does not endanger the AP1000 pressure boundary system.

Reliability Analysis of CVCS with Extended GO-FLOW Methodology by Considering Automatic Control Functions

Automatic Control System (ACS) has been more widely applied in nuclear power plants (NPP) in order to achieve better performance, convenience and economics. Some new challenges are also emerging in reliability analysis of ACS because of its complexity and strict regulatory requirements of NPP. The dynamic behavior of ACS is difficult to be expressed by traditional system reliability analysis techniques such as fault tree analysis and/or event tree which are widely applied for probabilistic safety assessment in NPPs. This paper presents an extended GO-FLOW methodology by introducing several supplementary operators for modelling control logic generally used in NPP systems. A Chemical and Volume Control System (CVCS) in Pressurized Water Reactor (PWR) NPP is taken as an example to demonstrate the usage of improved GO-FLOW methodology in modeling and reliability estimation of a process system coupled with automatic control functions.

The research and test of the cavitation performance of first stage impeller of centrifugal charging pump in nuclear power stations

The safety production is the most important index for nuclear power development, and the CVCS (chemical and volume control system) occupies an important place in safety control system. The CCP (centrifugal charging pump) is an important component of CVCS system and its particularity decides the high requirements on the cavitation performance, so the research on CCP cavitation has great significance. Using the 3D Pro/E software, the paper makes models of the flow passage components of first stage prototype of CCP in nuclear power stations, and completes the cavitation steady-state computation in the case that the high flow rate of first stage prototype reaches Q=160m3/h using the numerical computation software ANSYS CFX. The cavitation is divided into the following five stages according to its development process: inception operating condition, development operating condition, critical operating condition, serious operating condition and fracture operating condition. The research analyze the flow field, static pressure fields and bubble distribution conditions in the impeller in different cavitation states, and selects the middle flow line, the front shroud flow line and the back shroud flow line from the same blade for the comparative analysis; to verify the reliability of computation, the researcher made a first stage cavitation prototype to verify its hydraulic performance and cavitation performance. After the comparative analysis on the flow field, pressure and bubble distribution conditions of CCP first stage impeller in different cavitation states, the research draws the following conclusions: (1) as the inlet pressure decreases, the cavitation degree increases, the bubble distribution area extends from impeller inlet to impeller outlet and shows different laws in different cavitation states; (2) the generation of cavitation is affected by the pressure changes in impeller passages, and when the inlet pressure decreases, the static pressure in impeller passage decreases, thus producing a local low-pressure area; (3) the comparison of computation and test results shows small errors, thus proving the accuracy of analog computation; besides, the research verifies the cavitation performance in different flow rates by test, and test results show that NPSH3 is smaller than the 7.8m stipulated in Technical Specifications in any flow rate condition, indicating that the design is reasonable. The research points out a direction for the further improvement of optimization design of CCP cavitation performance.

Finalization of the conceptual design of the auxiliary circuits for the European test blanket systems

In view of the ITER conceptual design review, the design of the ancillary systems of the European test blanket systems presented in [1] has been updated and made consistent with the ITER requirements for the present design phase. Europe is developing two concepts of TBM, the helium cooled lithium lead (HCLL) and the helium cooled pebble bed (HCPB) one, having in common the cooling media, pressurized helium at 8MPa [2]. TBS, namely helium cooling system (HCS), coolant purification system (CPS), lead lithium loop and tritium extraction/removal system (TES–TRS) have the purpose to cool down the TBM and to remove tritium to be driven to TEP from breeder and coolant. These systems are placed in port cell 16 (PC#16), chemical and volume control system (CVCS) area and tritium building. Starting from the pre-conceptual design developed in the past, more mature technical interfaces with the ITER facility have been consolidated and iterative design activities were performed to comply with design requirements/specifications requested by IO to conclude the conceptual design phase.In this paper the present status of design of the TBS is presented together with the preliminary integration in ITER areas.

Thermal hydraulic feasibility analysis of the IBED PHTS for ITER

One of the main challenges of the ITER fusion reactor is to effectively remove large amount of heat deposited to the surface of the plasma facing components. The tokamak cooling water system (TCWS) will accomplish the objective of removing about 1GW of peak heat load from in-vessel components while maintaining pressures and temperatures of the coolant within acceptable and safe limits during different operational scenarios. A study of feasibility has been launched for the IBED PHTS (Integrated Blanket, Edge localized mode coils (ELMs) and Divertor Primary Heat Transfer System; it consists of five independent cooling trains (four operational and one in stand-by), one steam pressurizer, supply and return headers, ring manifolds and connections to the all in-vessel components (i.e. First Wall Blanket, Divertor, ELM, Diagnostics and other Ports clients).The dynamic behaviour of the IBED PHTS has been investigated by means of RELAP5® code to simulate the response of the system during plasma pulse and baking operations. Due to the plasma heat deposition on the surfaces of the in-vessel components and subsequent increase in hot leg temperature, a large amount of water volume is transferred from the hot legs of the circuit to the surge-line of the pressurizer during each burn cycle. This causes rapid increase of pressure and temperature of the system and the following actions are proposed to counteract these variations: spray injection in the upper dome of the pressurizer from the Chemical and Volume Control System (CVCS) to reduce the pressure and active control of flow rates through heat exchangers and their bypass loops to regulate the heat transfer from the primary system to the environment via secondary and tertiary loops.This paper focuses on the prediction of the thermal hydraulic behaviour of the IBED PHTS during plasma pulses and baking scenarios, describing the various activity of the analysis, the geometrical assessment of the circuit and the modelling with RELAP5® code. The results have been compared with design and operational requirement. Possible strategies to enhance the system performances have been formulated.

Numerical simulation of thermal stratification in an elbow branch pipe of a tee junction with and without leakage

Thermal stratification can cause thermal fatigue of the piping system of a nuclear power plant. One of the regions most at risk of suffering from thermal fatigue is a small elbow pipe branching off from the main pipe of the coolant loop for the drain or letdown system in the chemical and volume control system (CVCS). This work focuses on a fundamental description of the thermal stratification caused by turbulent penetration and buoyancy effects in the elbow branch pipe using large-eddy simulations (LESs). The LES results for the normalized temperature, mean temperature, and root-mean square (RMS) temperature were found to be in good agreement with the available experimental data which confirms that LES can predict the thermal stratification in a closed elbow branch pipe where cold fluids are stagnant. Subsequently, the flow and heat transfer were numerically predicted using LES when leakage occurred in the elbow branch pipe. The numerical results show that the thermal stratification region is pushed towards the horizontal part and may remain there for a long time for a leakage ratio of 1%. However, thermal stratification is quickly eliminated for a leakage ratio of 5%, although there is a higher power spectrum density (PSD) of the temperature in the early stages of the leakage. It may be concluded that a small leakage ratio can result in the elbow branch pipe being at high risk of thermal fatigue caused by thermal stratification.

Development of a boron concentration prediction model using multi-cell simulation of the automatic load follow operation

Load follow operations have been performed by manually changing the boron concentration in the reactor core and moving Control Element Assemblies (CEAs) for controlling the power and the power distribution. The manual operation of load follow requires experience and predictions related to core behavior following power changes because CEA movements distort the power distribution and a boron concentration control is also inaccurate and difficult due to the long time delay in boration/dilution operations. A boron concentration prediction model, accurately predict boron concentration in the reactor coolant system, including the chemical and volume control system (CVCS) following boration or dilution, was developed in order to enhance the boron concentration control during load follow operation. The model was developed using a multi-cell concept and integrated with the KOPEC Integrated Systems Performance Analysis Code (KISPAC), which is a system code used for design purposes. Boron concentration behavior was analyzed to verify the model for both direct and indirect injection using SKN 3&4 data. The load follow operation was simulated and the results were compared with the measured data obtained during the startup period. The developed model accurately predicted boron concentration behavior for all subsystems in the reactor control system and CVCS.

Investigations of Mixing Zones subject to Thermal Fatigue

Further the through wall crack in an elbow of the Residual Heat Removal (RHR) system of Civaux 1 in 1998, a large investigation program has been completed on removed parts of pipes to better assess the factors responsible for thermal fatigue cracking in pipe components where hot and cold fluids mix together. This article presents the results of destructive investigations performed on numerous parts of the RHR system of EDF plants, on a nozzle of the primary system and on a nozzle of the secondary system and their contribution to understanding and assessing harmfulness of the phenomenon.The damage concerns only certain RHR mixing zones: the mixing tee at intersection between downstream the exchanger and the by-pass line and the mixing tee at location of intersection of two portions of minimum flow line. Deep cracks are always associated with high temperature differences and long times of operation. The deepest cracks are always located close to the welds, in areas showing geometrical singularities and minimal thickness. Crack networks in running parts concern only the very first millimeters of the inner wall.Other mixing zones have been investigated: destructive examinations of a charging line primary nozzle of the Chemical and Volume Control System (CVCS), of two tees between the main line and the minimum flow line of the RHR system and of one tee connecting the Feed Flow Control System (FFCS) to the Auxiliary Feedwater System (AFS) did not show any thermal fatigue crack.EDF has developed a method for identification of fluid mixing zones subject to thermal fatigue damage. This method is based on the assessment of a usage factor, using the knowledge acquired during a large R&D program. This program investigated the thermal loadings in mixing tees acquired through several thermohydraulic mockup tests and the austenitic steel behavior under high cycle thermo-mechanical loadings.The results of the destructive examinations have been used to strengthen the identification methodology of mixing zones potentially affected by thermal fatigue. Damage found in the destructive examinations are always correlated with usage factors a lot greater than 1.The methodology has been applied to all mixing zones of safety related circuits of plants operated by EDF. It has lead to identify the ones with a usage factor greater than unity and to an appropriate in-service inspection program based on volumic non-destructive examinations.

Socket Weld Integrity in Nuclear Piping under Fatigue Loading Condition

The purpose of this paper is to evaluate the integrity of socket weld in nuclear piping under the fatigue loading. The integrity of socket weld is regarded as a safety concern in nuclear power plants because many failures have been world-widely reported in the socket weld. Recently, socket weld failures in the chemical and volume control system (CVCS) and the primary sampling system (PSS) were reported in Korean nuclear power plants. The root causes of socket weld failures were known as the fatigue due to the pressure and/or temperature loading transients and the vibration during the plant operation. The ASME Boiler and Pressure Vessel (B & PV) Code Sec. III requires 1/16 inch gap between the pipe and fitting in the socket weld with the weld leg size of 1.09*t 1 , where t 1 is the pipe wall thickness. Many failure cases, however, showed that the gap requirement was not satisfied. In this paper, the socket weld integrity under the fatigue loading was evaluated using three-dimensional finite element analysis considering the requirements in the ASME Code. Three types of loading conditions such as the deflection due to vibration, the pressure transient ranging from P=0 to 15.51 MPa, and the thermal transient ranging from T=25 °C to 288 °C were considered. The results are as follows; (1) The socket weld is susceptible to the vibration where the vibration levels exceed the requirement in the ASME Operation and Maintenance (OM) Code [9]. (2) The effect of pressure or temperature transient load on socket weld in CVCS and PSS is not significant owing to the low frequency of transient during plant operation. (3) 'No gap' is very risky to the socket weld integrity for the systems having the vibration condition to exceed the requirement specified in the ASME OM Code and/or the transient loading condition from P=0 and T=25 °C to P=15.51 MPa and T=288 °C.

Socket weld integrity in nuclear piping under fatigue loading condition

The purpose of this paper is to evaluate the integrity of socket weld in nuclear piping under the fatigue loading. The integrity of socket weld is regarded as a safety concern in nuclear power plants because many failures have been world-widely reported in the socket weld. Recently, socket weld failures in the chemical and volume control system (CVCS) and the primary sampling system (PSS) were reported in Korean nuclear power plants. The root causes of the socket weld failures were known as the fatigue due to the pressure and/or temperature loading transients and the vibration during the plant operation. The ASME boiler and pressure vessel (B & PV) Code Sec. III requires 1/16 in. gap between the pipe and fitting in the socket weld with the weld leg size of 1.09 × t 1, where t 1 is the pipe wall thickness. Many failure cases, however, showed that the gap requirement was not satisfied. In addition, industry has demanded the reduction of weld leg size from 1.09 × t 1 to 0.75 × t 1. In this paper, the socket weld integrity under the fatigue loading was evaluated using three-dimensional finite element analysis considering the requirements in the ASME Code. Three types of loading conditions such as the deflection due to vibration, the pressure transient ranging from P = 0 to 15.51 MPa, and the thermal transient ranging from T = 25 to 288 °C were considered. The results are as follows; (1) the socket weld is susceptible to the vibration where the vibration levels exceed the requirement in the ASME operation and maintenance (OM) code. (2) The effect of pressure or temperature transient load on socket weld in CVCS and PSS is not significant owing to the low frequency of transient during plant operation. (3) ‘No gap’ is very risky to the socket weld integrity for the systems having the vibration condition to exceed the requirement specified in the ASME OM Code and/or the transient loading condition from P = 0 and T = 25 °C to P = 15.51 MPa and T = 288 °C. (4) The reduction of the weld leg size from 1.09 × t 1 to 0.75 × t 1 may induce detrimental effect on the socket weld integrity.

Modeling the activity of 129I and 137Cs in the primary coolant and CVCS resin of an operating PWR

Mathematical models have been developed to describe the activities of 129I and 137Cs in the primary coolant and resin of the chemical and volume control system (CVCS) during constant power operation in a pressurized water reactor (PWR). The models, which account for the source releases from defective fuel rod(s) and tramp uranium, rely on the contribution of CVCS resin and boron recovery system as a removal process, and differences in behavior for each nuclide. The current models were validated through measured coolant activities of 137Cs. The resultant scaling factors agree reasonably well with the results of the test resin of the coolant and the actual resins from the PWRs of other countries.

Modeling the activity and scaling factor of 129I in the primary coolant and CVCS resin of operating PWR

To estimate the activity of 129I at the primary coolant and chemical and volume control system (CVCS) resin in Korean pressurized water reactor (PWR) plants, a theoretical methodology was developed on the basis of an existing model of primary coolant activity and new model of CVCS resin activity. In order to reflect the difference between 129I and 137Cs, the different power-related diffusivities in the defective fuel were derived, and the variable removal efficiency of the CVCS resin for 137Cs was applied as a function of the coolant activity ratio of 131I/ 137Cs. The current computational method was validated by using the measured coolant activities of 137Cs, and the results show better agreement than a previously suggested parameter correlation method between 129I and 137Cs. There was also reasonable agreement in a comparison of the results of the test resin columns of the coolant from the PWR plants of other countries. It was shown that the ratio of the effective removal efficiency of 129I and 137Cs in the CVCS resin linearly influences the activity ratio of 129I/ 137Cs in the coolant, but on the other hand, its influence on the activity ratio in the CVCS resin is relatively less sensitive compared with that in the coolant.

Experiments on the thermal stratification in the branch of NPP

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth. the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems. establish the thermal stratification criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.

Numerical prediction of flow and mixing characteristics in CVCS chemical addition system

A numerical result of the flow and mixing characteristics is presented for the flow field created by water injected into a cylindrical tank with an initially stationary fluid. The flow is relevant to the operation of the chemical addition system in the chemical and volume control system (CVCS) of nuclear power plants. This study was undertaken to provide a basis for modification of the previous design which gave a number of difficulties in installation and operation of the chemical addition system because it needs a special reciprocating pump with a high actual head. For the tank of length-to-diameter ratios ( L/ D) of 1, 2 and 3, each with and without a baffle inside, calculation results were obtained by solving the unsteady laminar two-dimensional elliptic forms of governing equations for the mass, momentum and species concentration. Finite-difference method was used to obtain discretization equations, and the SIMPLER solution algorithm was employed for the calculation procedure. Results showed that the baffle was very effective in enhancing the mixing in the tank, and that a baffle should be installed near the tank entrance in order to inject chemicals into the reactor coolant system (RCS) within the operating time required.

A Concept of Passive Safety, Pressurized Water Reactor System with Inherent Matching Nature of Core Heat Generation and Heat Removal

The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of “safety class” for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation.

A Concept of Passive Safety Pressurized Water Reactor System with Inherent Matching Nature of Core Heat Generation and Heat Removal.

The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of “safety class” for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation.

Quantitative evaluation on design simplification of the chemical and volume control system

Recently, most of the countries using nuclear power for electricity generation are working towards the safer and the more economical designs. In practice some typical nuclear steam supply system (NSSS) vendors are developing evolutionary reactors or passive type reactors based on current technologies. One of the most important features of these advanced reactors is system simplicity. In this study, a model has been presented to quantitatively evaluate system simplification. Design enhancements in simplicity of the chemical and volume control system (CVCS) of the ABB-CE System 80+ and the Korean standardized nuclear power plant (KSNPP) have been evaluated using the presented model. The quantification model is based on the informational entropy concept and system availability analysis. The quantification results indicate that the simplicity of the CVCS is highly dependent on the type and the number of charging pumps. Also, it is found that higher redundancy does not always result in higher availability. Finally, it is stated that the informational entropy concept and the system availability analysis can be a valuable tool for the quantitative evaluation of system simplification.