Pump Trip Research Articles

A theory and analysis of the impact of gas in the dynamical behavior of the molten salt research reactor leading to the computation of the "gas coefficients of reactivity"

Background Molten salt reactors, and other types of circulating, liquid fueled, nuclear reactors contain a certain amount of gas entrained in their liquid nuclear fuel. This gas induces an effect on the nuclear and dynamical behavior of the reactor as a whole. Gas voids respond to variation in temperature and pressure differently than liquids. When the gas voids in the reactor working fluid expand, the nuclear fuel is pushed from the core. Likewise, when the gas voids contract, more nuclear fuel enters into the core. Methods This paper examines the interplay of gas void fraction and reactivity in a molten salt reactor, and attempts to elucidate the dynamical response of the void fraction and the reactivity of the system to perturbation in system temperature, pressure, and gas quantity. A theory is presented that aims at describing the relationship between reactivity and gas behavior. This theory is then applied to the Molten Salt Research Reactor (MSRR) design, a facility currently under construction at Abilene Christian University campus. Results A result of this paper is the temperature and void fraction parameterized gas coefficients of reactivity for the Molten Salt Research Reactor. Conclusions The presence of voids accounts for 5-30% of the total temperature coefficient of reactivity, demonstrating their non-trivial contribution. Additionally, the study emphasizes the importance of considering gas content in MSR physics, especially in the context of pressure transients and system reactivity during pump trips. The initial system pressure, particularly in designs like the MSRR operating at sub-atmospheric pressures, is crucial due to its influence on reactivity changes during rapid pressure increases.

Multi-physics code coupling methodology for the simulation of accidental transients in sodium cooled fast reactors

The French Alternative Energies and Atomic Energy Commission (CEA) has played a substantial role in advancing fourth-generation reactor technologies, including Sodium-cooled Fast Reactors (SFR), which are anticipated to offer enhanced efficiency, adaptability, and sustainability compared to existing nuclear technologies. In the context of safety assessment, it is crucial to precisely simulate accidental scenarios like Unprotected Loss Of Flow (ULOF) transients. These transients entail intricate multi-physics interactions within the reactor, particularly within the core, encompassing thermal-hydraulics, reactor physics, and fuel performance (e.g. changes in densities or temperatures will lead to neutronics effects, thus power, then causing feedbacks onto said densities and temperatures).In order to capture those complex phenomena, a multi-physics coupling approach for SFRs, which combines the CATHARE3 code for system scale thermal-hydraulics, the APOLLO3 code for reactor physics, and the GERMINAL V2 code for nuclear fuel performance has been developed at CEA in the recent years.This paper presents a comprehensive methodology of such coupling and its integration into a Verification, Validation and Uncertainty Quantification (VVUQ) process. The first part of the paper provides an introduction to the simulation tools used for the coupling. The following section explains the multi-physics coupling approach, facilitated by CEA’s Collaborative Code Coupling Platform (C3PO), which enables code integration through a Python interface to create a comprehensive simulation model. A detailed examination of the coupling algorithm and data transfer mechanisms is also provided.The generality and reliability of this approach are demonstrated by simulating two accidental transients: the Unprotected Loss Of Flow (ULOF) in the ASTRID reactor and the Loss Of Flow WithOut Scram (LOFWOS) in the Fast Flux Test Facility (FFTF), both corresponding to a primary pump trip without control rods falling. The results show that the proposed multi-physics coupling scheme is effective and robust for SFR simulations. The coupled model enables a more accurate and realistic representation of the complex phenomena occurring in the reactor during an accident.This research underscores the significance of multi-physics coupling in the analysis of SFRs and serves as a valuable point of reference for future SFR studies. The proposed approach offers an efficient way to investigate SFR responses during accidents or incidents, thereby advancing the development of dependable and precise simulation tools for SFR design and evaluation.

Negative pressure protection with one-way surge tanks in bidirectional water delivery systems

ABSTRACT Water hammer protection is one of the key factors for the safe operation of the long-distance water delivery systems, which has lots of useful results. However, the protection experiences for bidirectional water delivery systems using the same pipeline are lacking. In this study, the protection method with one-way surge tanks in the bidirectional water delivery system is investigated. The theoretical formula for one-way surge tank setting is derived based on the theory of water hammer propagation, and the pump head and bidirectional water hammer protection constraints. Subsequently, the pump trip transient processes in an actual system are simulated. Finally, different kinds of one-way surge tank protection schemes are compared. The results show that the pressure along the pipeline after the pump trip can meet the standard when using the one-way surge tank protection scheme calculated by the theoretical formula. The proposed method can provide the effective protection strategy for the bidirectional water delivery systems.

Exploring the Sensitivity of the Transient Response following Power Failure to Air Valve and Pipeline Characteristics

Air valves are protective devices often used in pressurised water pipelines, ideally admitting air to limit sub-atmospheric pressures and controlling the release of entrapped air. This work summarises a comprehensive sensitivity analysis of the transient behaviour in a rising water pipeline with an air valve following a pump trip. The paper examines the water hammer stages associated with a pump trip, namely, the initial depressurisation, followed by air admission, then air expulsion, and finally the creation of a secondary pressure wave. For each air valve location and specific set of design conditions, the relationship between the transient magnitude and air valve outflow capacity is found to be non-linear, but to roughly follow the shape of a logistic curve having a lower left plateau for attenuated (type 1) behaviour and transitioning through type 2 behaviour to a higher right plateau for water-hammer-dominated (type 3) behaviour. Through an extensive set of simulations covering a wide range of conditions, the study identifies the size of the critical outflow orifices associated with both type 1 and type 3 responses and assesses the influence of the location of the air valve on the transient magnitude and on the timing of air pocket collapse. Furthermore, the paper highlights that a non-slam air valve is capable of effectively mitigating transient magnitudes provided that its design parameters are judiciously chosen and account for both the system’s attributes and the characteristics of the transient event.

Study on start-up of molten salt fast modular reactor from shutdown based on two-phase flow CFD analysis

In the present study, the effect of helium gas bubbles injected into a molten salt reactor on nuclear characteristics is analyzed. The analysis uses a computational fluid dynamics code and a system code to evaluate the effects of void fraction and core temperature on reactor power, which are affected by changes in fuel pump speed. Because molten salt reactors are equipped with a fission gas removal system such as Xe gas, the core of a molten salt fast reactor is strictly a two-phase flow and the analysis considers the effects of voids. In this system, the Xe gas is removed by injecting helium gas bubbles into the molten fuel salt. When the void fraction of the helium gas increases by 1%, the reactivity of approximately −360 pcm δk/k is introduced. To understand the importance of two-phase flow analysis by the FLUENT code with a user defined function which incorporates the point kinetics model and void effect model, a fuel pump trip at rated reactor power with a void fraction of 0.4% in the core average was analyzed with FLUENT and compared with a conventional analysis that does not consider the effects of voids at all. The results show that the transient behavior cannot be properly evaluated without considering the effects of voids. A comparison of two-phase and single-phase flow calculations using FLUENT is also provided, and a method to evaluate the effects of voids in single-phase flow calculations is also proposed. Analysis using FLUENT under two-phase flow conditions shows that the reactor can increase power by increasing the fuel pump speed. Under single-phase flow assumption, the analysis was performed using RELAP5-3D and showed that the reactor can be started in a similar manner by considering the reactivity associated with changes in the void fraction. It was pointed out that the reactor could be controlled by adjusting the amount of helium injected. It was also noted that voids increase the reactor shutdown margin.

Validation of the Ants-TRACE code system with VVER-1000 coolant transient benchmarks

The objective of this paper is to start the validation of the coupled code system of the nodal neutronics solver Ants and the system code TRACE. The process consists of two exercises of well-known VVER-1000 coolant transient benchmarks: a core-vessel simulation of a main steam line break transient of the V1000CT-2 benchmark and a full plant simulation of a main coolant pump trip of the Kalinin-3 benchmark. The Ants-TRACE results are compared against other code solutions in the V1000CT-2 benchmark and measurement data in the Kalinin-3 benchmark. The results show good agreement in both benchmarks, with the deviations ranging mostly within the measurement error or the range of deviations of other published solutions. The results confirm the correct implementation of the coupling and successful modeling of relevant phenomena in coolant transient events in a VVER-1000 reactor.

Safety-oriented SFR core design: Methodology and application to transient bifurcations and stability maps analyses

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Alternative Energies and Atomic Energy Commission), a methodology is proposed to early consider safety requirement in the undergoing reactor design process. Before the use of mechanistic tools (CATHARE, SIMMER, EUROPLEXUS, etc.) whose input deck elaboration requires an advanced knowledge of the reactor design, the methodology proposed in this article involves several physical tools simulating phenomena likely to govern the choice of design parameters. These tools are mostly based on reduced order models (ROM), meaning that they mainly involve low-dimensional modelings (mostly 0D and 1D) that are validated versus experimental results. They are gathered in a platform that covers all kind of accidental phenomenology, from the initiator (pump trip, reactivity insertion, local flow blockages, etc.) until the reach of a stable and coolable state after corium relocation. Thus, enabling a large number of simulations in a reasonable computational time, this fast-running platform makes possible the characterization of some major accident transient bifurcations (such as boiling onset, boiling stabilization, primary power excursion, molten fuel vaporization, corium axial relocation in transfer tubes, etc.), in terms of probability of occurrence and of consequences on the transient evolution. It also enables to identify the main physical parameters causing the bifurcations in order to allow a straight feedback on the core design and to give some orientations for future R&D studies. In this article, a focus is firstly made on some scenario bifurcations to illustrate the platform capabilities. The boiling onset and possible reactor state stabilization are studied. The possibility of primary power excursion depending on the core design and the fuel vaporization possibilities are assessed. This platform is also of great help in order to rapidly compare several core designs when facing accidental transients. An example is given by comparing a CADOR core design to an ASTRID-like core design through flow stability maps in natural circulation conditions obtained by the fast-running tool platform. These applications demonstrate the efficiency of the presented methodology integrating safety at the very first design stages, and at facilitating the safety-oriented design of SFRs.

Loss of flow accident analysis in Atucha II nuclear power plant using RELAP5 model

The loss of flow accident (LOFA) in nuclear power plants (NPP) is considered a design base initiating event (DBIE) and it is broadly analyzed in safety and licensing evaluation. Therefore, this is commonly chosen to assess computational models. The current paper addresses the simulation of LOFA under single and total main coolant pump (PMP) trips during full power in Atucha-II Pressurized Heavy Water Reactor (PHWR). RELAP5-mod3.3 NPP was used for solving the thermal-hydraulics, neutron kinetics, and control and safety logic in steady-state and LOFA events. Simulations were assessed against real plant data recorded during a single-PMP trip obtaining very good agreement. After that, the total-PMP trip was simulated, showing that the plant also reached a safe steady condition, guaranteeing the removal of decay heat by natural flow circulation. The very good agreement in terms of the main thermal-hydraulics parameters demonstrated the capability of the computational model to reproduce the overall NPP behavior.

Design of the Cooling Water System for the EAST Lower Divertor

The new lower tungsten divertor is a crucial component to achieve the targets of long pulse and high <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\beta $ </tex-math></inline-formula> H-mode operation of Experimental Advanced Superconducting Tokamak (EAST) device. The cooling water system (CWS) of lower divertor should have sufficient heat removal capability and be able to cope with pressure transients caused by the potential water hammer events, which makes the design challenging. In this study, the system process and details of main component are determined based on the design requirements. The thermal-hydraulic analysis model has been developed by Flowmaster (FLOWMASTER) to conduct the heat transfer analysis. Furthermore, the system water hammer transients under one pump trip condition have been analyzed. The results show that the flow distribution of each client is uniform, and the pressure and temperature of key points meet the design requirements. The maximum pressure fluctuation, 0.16 MPa, caused by the sudden closing of the check valve is within the acceptable range. The experiment data are in good agreement with the simulation values, which proves the correctness of the model and the feasibility of the design.

Analysis and optimization of operation mode of outlet communication valves for circulating pump in cold season

The circulating cooling water system of a coastal power plant adopts a unit seawater supply system. Each unit is arranged with three circulating water pumps. The operating mode of the circulating water pump in the cold season is two for operation and one on standby. If a running circulating water pump trips due to a fault, the standby pump will be activated by interlocking; meanwhile, an opening or closing command will be issued to the two outlet communication valves. According to the state of the outlet communication valve, the risks such as overpressure, condenser vacuum, water hammer, and pump reversal are analyzed. In order to reduce the possible risks in the cold season, an optimization scheme is proposed for keeping the outlet communication valves fully open. The feasibility of the optimization scheme in terms of flow distribution and pipe and equipment stability is also analyzed. The outlet communication valve optimization scheme will reduce operational risk and ensure the safe and stable operation of the circulating water system. The optimization scheme provides reference and experience feedback for the follow-up circulating water system design and operation.

The Crucial Importance of Air Valve Characterization to the Transient Response of Pipeline Systems

Air valves are often crucial components in an air management strategy for pressurized water conveyance systems. However, the reliability of characteristic curves of air valves found in product catalogs is quite variable. This paper evaluates the consistency of a selection of product curves to basic air flow principles. Several recurring issues are identified: catalogs that present identical curves for admission and expulsion (they are, in fact, quite distinct); admission curves that are inconsistent with the isentropic inflow model; inflow (admission) curves actually consistent with the shape of the isentropic outflow model; limited validity curves that encompass only part of the subsonic flow regimen; and unclear or unstated specifications regarding the conditions under which the characterization tests were performed or their results displayed. To examine the significance of these representational issues related to air valve capacity on system behaviour, this paper uses a case study involving the simulated transient response arising from a pump trip at the upstream end of a rising water line having a distinct high point fitted with an air valve. It is found that employing inaccurate air valve characteristics in a transient simulation may potentially result in appreciable or even dangerous simulation errors.

Concurrent 6. Presentation for: Effects of rapid gas decompression on coal seam gas swellables

Presented on Tuesday 17 May: Session 6 Thousands of non-stimulated coal seam gas (CSG) wells in Queensland’s Surat Basin rely on swellable packers as the first line of defence against interburden solids production and poor well run life. This paper is aimed at understanding some of the impacts of long-term well operations on swellable performance from rapid changes in downhole pressure. For the first time, rapid gas decompression (RGD) effects on CSG swellables were experimented on in a quantitative manner as an analogue to underbalanced workover and pump trip conditions. RGD has the potential to break down swellables due to rapid release of high-pressure methane diffused in the rubber matrix resulting in a flow path for interburden solids. Five commonly available swellables from the CSG market were lab-tested for rapid decompression with methane at operational conditions. Coupon samples were swollen to representative conditions and placed in an autoclave under high-pressure methane, then rapidly decompressed in cycles. Results of this study showed relatively low levels of physical degradation under test conditions but shrinkage effects varied between products largely grouped into material properties, confirmed with separate ambient shrinkage test. As such, the focus on swellable placement geometry remains paramount. To access the presentation click the link on the right. To read the full paper click here

Instantaneous acceleration-based modeling of pumping systems response under transient events

Water hammer can cause pipeline eruptions after uncontrolled pump trip. In order to protect the pipeline from water hammer effects and monitor the transient flow, it is imperative to accurately predict pressure oscillation and surge energy dissipation for longer simulation periods. This study investigates the effectiveness of the water hammer control strategy by adjusting the rotational inertia of a pumping system alongside valve closing operations. In addition, it also explores the complex nature of the transient energy dissipation by using the one-dimensional instantaneous acceleration-based (IAB) unsteady friction models. To this end, the IAB unsteady friction (UF) models were improved to calculate the water hammer pressure by investigating the elastic pipe's response under transient events that are caused by pump trip and valve closure. Various boundary equations were derived and combined with water hammer governing equations to form the final system of equations to predict the transient pressure variation with time. After that, the numerical results of two IAB friction models (i.e., one and two-decay coefficient models, respectively) were compared with previous experimental data to verify the proposed transient solver. This solver was based on Method of Characteristics (MOC) and UF models. The results indicate that one-coefficient and two-coefficients UF models yield almost similar results that match with the experimental outcomes; however, the latter model is comparatively more complex in terms of formulation. Furthermore, a parametric study shows that increasing the moment of inertia of the pump and integrating it with the optimal operation of the valve helps in significantly controlling the transient pressure. This helps in reducing the check-valve slam and controls the water hammer with no need for additional protection devices. Results of the integral energy equation describe that the role of UF on the total dissipation of transient energy diminishes with an increase in the valve's closing time and inertia of the pumping system. Finally, the IAB-UF model acts as a valuable numerical tool that provides comprehensive information about the physical parameters required to protect high-head water supply systems from transient incidents.

Effects of rapid gas decompression on coal seam gas swellables

Thousands of non-stimulated coal seam gas (CSG) wells in Queensland’s Surat Basin rely on swellable packers as the first line of defence against interburden solids production and poor well run life. This paper is aimed at understanding some of the impacts of long-term well operations on swellable performance from rapid changes in downhole pressure. For the first time, rapid gas decompression (RGD) effects on CSG swellables were experimented on in a quantitative manner as an analogue to underbalanced workover and pump trip conditions. RGD has the potential to break down swellables due to rapid release of high-pressure methane diffused in the rubber matrix resulting in a flow path for interburden solids. Five commonly available swellables from the CSG market were lab-tested for rapid decompression with methane at operational conditions. Coupon samples were swollen to representative conditions and placed in an autoclave under high-pressure methane, then rapidly decompressed in cycles. Results of this study showed relatively low levels of physical degradation under test conditions but shrinkage effects varied between products largely grouped into material properties, confirmed with separate ambient shrinkage test. As such, the focus on swellable placement geometry remains paramount.

Study on water hammer protection of the siphon breaking structure in the water supply system

Abstract An appropriate water hammer protective scheme is a significant concern in the operation of water supply projects. According to the special terrain in the water supply project, which forms a siphon breaking structure at the end of the pipeline, three protective schemes were proposed and compared: single vacuum breaking valve (VBV) scheme, VBV and air valve scheme, and VBV and one-way surge tower scheme. Based on the control standards of pipe pressure, the three protective schemes were assessed in terms of suppressing the negative pressure caused by a pump trip accident. The results show that the siphon breaking structure with the VBV can achieve good effect protection only in a limited range of pipelines. In the VBV and air valve scheme, the pressure oscillations were obviously caused by repeated inlet and exhaust of the air valves. To avoid supplementing too much gas in the pipe by air valves, which will result in a gas column bridging phenomenon, the VBV and one-way surge tower scheme is proposed and can better meet the requirement of the pressure control standard.

Vibration analysis for predictive maintenance and improved reliability of rotating machines in ETRR-2 research reactor

Abstract In this work, both hardware and software modifications in a typical research reactor protection system (RPS) is proposed. The reactor cooling pumps are tripped based on vibrations safety signals of the pumps while the reactor SCRAM signal is initiated based on low flow rate and pressure drop across the reactor core which is a direct result of pumps trip. The main objective of this work is to develop reactor SCRAM signal based of core cooling pumps vibration signals. The early shutdown of the reactor based on pumps vibration signals is of significant importance not only in cooling the decay power of the reactor core after shutdown but also to prevent pumps failure. In the hardware model, the core cooling pumps vibration signals are feed to RPS to initiate reactor SCRAM signal. In the software model, a modular artificial neural network (ANN) is used in modeling the vibration monitoring of the research reactor (ETRR-2). The input and the output signals of the vibration transducer are used as a source data for training the neural network model. The type of the network used in this methodology is the supervised Multilayer Feed-Forward Neural Networks with the back-propagation (BP) algorithm. Vibration analysis programs are used in research reactors (RRs) to identify faults in machinery, plan machinery repairs, and keep machinery functioning for as long as possible without failure. The vibration severity limits are determined based on the International Organization for Standardization (ISO) 10816. The ANNs were designed using two different methods; one is by using hardware application contained two out of three voting and dynamic modules for trip signal by using ANNs. The current model classifies the vibration signals into five ranges low, good, satisfactory, unsatisfactory, and unacceptable vibration. The ANN is trained to detect the signal and vote to take the correct and safe action. The results demonstrate that the ANN can help in taking predictive actions for the safe core coolant pumps operation.

Analyzing Simplified BWR Inherent Safety System using IAEA Generic Boiling Water Reactor Simulator

The inherent safety of boiling water reactors (BWR) has been a vital research topic in the past decades. This study aimed to observe and analyze the simplified BWR inherent safety system incorporated in IAEA Generic BWR Simulator. This simulator represents important features of BWR and provides graphical information and real-time simulation data. The simulated BWR has 1300 MWe power with ABWR-type containment. To analyze its inherent safety system, three conditions are simulated, i.e., normal condition at 100% power, transient condition (feedwater pumps trip), and emergency condition (loss of coolant accident—LOCA). The simulations were performed for up to 30 minutes since the most critical events in all conditions occurred within that time frame. Sequences of transient and emergency conditions were described in detail with the help of an additional screen recorder and time counting software. Results of several parameters in all simulation conditions were compared and analyzed. It was concluded that the simulator could simulate the normal, transient, and emergency conditions and the simplified version of the BWR inherent safety system.

Numerical simulation of transient overflow characteristics in CCS pump building after the main feed water pipe of SG ruptures

The feed water pipes of steam generators (SG) stretch across the pump building of the component cooling water system (CCS), thereby connecting the reactor containment and the steam turbine building in the pressurized water reactor of a nuclear power plant. When main feed water pipes rupture, large quantities of water and steam may discharge into the CCS pump building, which causes the CCS pumps to be submerged and the CCS system to shut down, thereby threatening the safety of nuclear power plants. Therefore, it is necessary to analyze the overflow characteristics after such an accident. However, it is difficult to perform experiments in the CCS pump building, owing to its huge area and complex structure. In this study, numerical simulation was performed via computational fluid dynamics by using the volume of fluid model, in which three phases (water, air, and steam) were considered. Validation & Verification (V&V) is performed by comparing the calculated results with experimental verification of the small model, so that the reliability of the calculation method and results are proved. Overflow characteristics in three different accidental conditions were also analyzed: the pump trip accident triggered by double-end fracture (TDEF), pump trip accident triggered by conservative maximum flow rate discharge (TmaxD), and pump trip accident not triggered by a fracture (TnotF) in the pressure compartment. It has been revealed that under TDEF and TmaxD, the maximum water level around the two CCS pumps is lower than the waterproof foundation bed of the CCS pumps. However, under TnotF, the highest water level around the two CCS pumps is much higher than the waterproof foundation bed of the CCS pumps, which may threaten their safety. To address this problem, certain new optimization design schemes have also been proposed.

CFD simulation of the pump trip runaway transient process of a pumped-storage power plant with head 700 m

Simulating transient processes of pumped-storage power plants is essential for the design and optimization of pump-turbine and water conveyance system. Because pressure pulsations and runner forces cannot be obtained from the traditional 1D transient simulation methods, 3D CFD simulations are necessary. In this study, the fast runaway transient process, occurring after the pump trip of a pumped-storage plant with a head of 700m, was simulated by using the CFD method that couples the 1D water conveyance system and 3D turbine. The phenomena, composed of water hammer fluctuations, pressure pulsation transitions, and flow pattern evolutions, were described and analyzed. We found that the flow patterns and pressure pulsations change violently as the working point goes through different regions. Especially, the pressure pulsations and unbalanced radial runner forces are the most severe in the S-shaped characteristic region. The strong impact on blades, large scale recirculating vortices in blade passages, obvious rotating stalls in vane and blade regions, transitional backflows at runner inlet and outlet, and strong circumferential velocity in vaneless space are the main sources. The dynamic working trajectory does not follow the measured static characteristic curve and demonstrates an undamped loop in the S-shaped region. This paper provides some evidence on the relations between flow structure and pressure pulsation features, although the mechanism of dynamic characteristics needs further investigations.

Numerical simulation on transient thermal and hydraulic characteristics in sodium pool of CEFR under OPT-SLOOP and OPT-RHROSL conditions

One Pump Trip (OPT) may cause asymmetric phenomena in China Experimental Fast Reactor (CEFR) primary system. Not only two loops asymmetric coast-down processes and failure pump backflow, but superposition loss of off-site power (SLOOP) or residual heat removal of single loop (RHROSL) with OPT may make the situation more complicated. In order to investigate transient thermal hydraulic characteristics in sodium pool under OPT-SLOOP and OPT-RHROSL conditions, a full-scale numerical 3D model of sodium pool in CEFR was established, in which the transient processes of coast-down and backflowwere simulated. The simulation results revealed that, in the case of OPT-SLOOP, the intact loop overall temperature was significantly higher than the failure loop overall temperature in the cold sodium pool, and the maximum IHX outlet temperature difference between the two loops reached 125.7℃, while in the case of OPT-RHROSL, the maximum IHX outlet temperature difference was limited to 15℃.