Minimum Critical Heat Flux Ratio Research Articles

Design and analysis of KA-130, the marine-based SMR for icebreakers

Designed for icebreakers, the KA-130 integral SMR provides a carbon-free solution for navigating the North Pole Route with nuclear propulsion. As an integral reactor, all primary components are integrated into the reactor pressure vessel, eliminating large-break loss-of-coolant accidents and enhancing reactor safety. The study evaluates its load following capability and demonstrates efficient load changes within two minutes. The safety analysis during small break loss-of-coolant accidents (SBLOCA) incorporates ship movements into the minimum critical heat flux ratio (MCHFR) calculations and includes a sensitivity analysis to identify significant scenarios. The most limiting case was the guillotine break at the safety injection line with low initial pressure, high core exit temperature and bottom-peaked core. The analysis shows that core flooding is maintained, and decay heat removal is successfully performed. Acceptance criteria, aligned with regulatory and KA-130 specific standards, ensure safety during and after SBLOCA, meeting MCHFR and collapsed water level conditions above the core.

Experimental and analytical natural circulation analysis of the Jordanian 5 MW research reactor using plate type fuel by RELAP5 MOD3.3

The 5-MWth Jordanian research reactor is an open-pool type research reactor using plate-type fuel. It is cooled by the upward natural flow at low power-operation. Natural circulation cooling is important in the safety analysis because it is the last scenario of all accidents. Therefore, studying natural circulation is important in determining the safety limits, such as the maximum allowed temperature and the minimum critical heat flux ratio of fuel. In this paper, the reactor operated at 50KW by the natural circulation to investigate its neutronics and thermal–hydraulic characteristics. Then, the results were benchmarked with RELAP5 MOD 3.3 by implementing models with and without considering the effect of heat loss through the Upper Guide Structure. As a result of this study, different models show minor differences in thermodynamic parameters and an additional safety margin. Eventually, benchmarking the results and the calculations of RELAP5 led to a reasonable agreement in both simulated models.

Steady-State Subchannel Thermalhydraulic Assessment of a Full-Scale Pressurized Water Reactor-Small Modular Reactor Fuel Assembly With Conventional and Advanced Fuels

Abstract Thermalhydraulic analyses using subchannel codes (e.g., ASSERT-PV) are performed as a support tool to evaluate safety margins and the key parameters. Advanced fuels have recently attracted the international community's interest to improve safety margins during normal operation and accident scenarios by utilizing special coatings and barriers in a heterogeneous, multiregion, multicoating, multiclad annular fuel. In addition, advanced fuels improve the performance characteristics such as a higher burnup and better uranium utilization. Therefore, there is a need to understand the implications of such advanced unconventional fuels for the landscape of the Canadian nuclear industry and Canadian policy for energy development. In this work, subchannel thermalhydraulic analysis of a small modular reactor based on pressurized water reactor technology (PWR-small modular reactor (SMR)) core is performed using ASSERT-PV. A benchmark of a 17 × 17 fuel assembly with conventional fuel elements in comparison to PWR-SMR specification was conducted. The advanced fuel element system is also investigated and compared with the conventional one. The results indicated that the advanced fuel achieves a significant reduction in fuel element temperature by 15%. In addition, the results revealed that the proposed advanced fuel could achieve a minimum critical heat flux ratio (MCHFR) higher than the conventional fuel by 17%.

Conservative analytical and experimental investigation of all pumps failure event in the primary cooling system for a 5 MWth research reactor

The 5-MWth Jordanian Nuclear Research Reactor is an open-tank and open-pool research reactor which uses narrow rectangular fuel. The reactor is cooled by demineralized water in downward forced flow and natural circulation in transient events. The reactor performed a loss of flow test in the Primary Cooling System (PCS) by turning off all PCS pumps at full-power operation. The reactor then displayed safe behavior in thermal-hydraulic, neutronics, and acceptable radiological values. Besides, the RELAP5 MOD3.3 code was used to implement the reactor in three different comprehensive models according to several safety analysis conservatisms, such as assuming a lower total core flow rate by −12% or assuming failure in a safety valve. After that, a benchmark was conducted between the experimental results, and the code calculations of implemented models showed well agreement in general thermal-hydraulic and safety analysis parameters. The code proved a safe reactor trend that used all conservative assumptions even in the most conservative implemented model. The primary safety parameters, the Maximum Temperature of Fuel Clad (MTFC) and the Minimum Critical Heat Flux Ratio (MCHFR), were 421 K and 1.82 for a short time, respectively. The resulting parameters underwent the event's acceptance criteria, where the limits of the set values are 673 K and 1.5 for MTFC and MCHFR, respectively. As a result of the study, several models illustrated differences in thermodynamic parameter effects and gave higher and additional reactor safety levels. Moreover, the test comprehensively investigated the reactor behaviors and ensured the reactor cooling system capability by free convection cooling during the event.

Thermal-hydraulic modelling and analysis of ITER tungsten divertor monoblock

• A mathematical model has been developed to investigate the thermal-hydraulic performance of ITER tungsten divertor monoblock. • The model predicts the thermal response of the divertor structural materials and coolant tube. • The model accounts for the melting, vaporization, and solidification of the upper layer of the divertor facing plasma. • The model is verified against a previous calculation in the literature for DEMO divertor at an incident surface heat flux of 10 MW/m 2 . • The model is used to simulate the divertor materials response subjected to a VDE. The divertor is a fundamental component of fusion power plants, being primarily responsible for power exhaust and impurity removal via guided plasma exhaust. Due to its position and functions, the divertor has to sustain very high heat flux arising from the plasma (up to 20 MW/m 2 ), while experiencing an intense nuclear deposited power, which could jeopardize its structure and limit its lifetime. Therefore, attention has to be paid to the thermal–hydraulic design of its cooling system. In this work a mathematical model has been developed to investigate the steady state and transient thermal–hydraulic performance of ITER tungsten divertor monoblock. The model predicts the thermal response of the divertor structural materials and coolant tube. The coolant tube is divided into specified axial regions and the divertor plate is divided into specified radial zones, and then a two-dimensional heat conduction calculation is performed to predict the temperature distribution for both steady and transient states. A two-dimensional numerical finite difference technique is adapted in Cartesian coordinate system where the implicit scheme is used for transient calculation. The model also accounts for the melting, vaporization, and re-solidification of the upper layer of the divertor facing plasma. The selected heat transfer correlations cover all possible operating conditions of ITER under both normal and off-normal situations. The model is verified against a previous calculation in the literature for DEMO divertor at an incident surface heat flux of 10 MW/m 2 . The model is then used to predict the steady state thermal behaviour of the divertor under incident surface heat fluxes ranges from 2 to 20 MW/m 2 for a bare cooling tube and cooling tube with swirl-tape insertion. It calculates the maximum tube surface heat flux and the minimum critical heat flux ratio for all cases as well. The model is also used to simulate the divertor materials response subjected to high heat flux during a vertical displacement event (VDE) where 60 MJ/m 2 plasma energy is deposited over 500 ms.

Analysis of a loss-of-flow accident resulting from the primary pump shaft break transient of the NUR research reactor

ABSTRACT This paper presents a combined experimental and numerical simulation approach to investigating safety issues related to postulated loss-of-flow accident (LOFA) cases, which are more likely to occur in the NUR Research Reactor (Algeria). The transients investigated at nominal-power operating conditions are related to the loss of flow resulting from an instantaneous shaft break in the main cooling pump of the NUR reactor. The investigations are based on hydrodynamic and thermal hydraulic experiments to assess the reactor cooling system’s behavior. 3D Monte Carlo neutron transport calculations were performed with the (MCNP) code to determine the resulting neutronic properties of the core. In the accident analysis, a model of the primary cooling system was applied via the RELAP5 code. The experimental data and RELAP5 predictions showed good agreement. Additionally, the LOFA due to the transient scenario of the pump shaft break was compared with the LOFA due to normal loss of the coolant pump power transient. The results obtained from the transient (LOFA) studies revealed that in both cases, the lower limit of the minimum critical heat flux ratio and minimum onset of flow instability ratio for NUR is satisfied with a sufficient margin.

Steady-State Subchannel Thermal-Hydraulic Assessment of Proposed Thorium Fuel Designs

The Canadian Nuclear Laboratories (CNL) is developing the technologies to enable the use of thorium-based fuels in pressure tube–heavy water reactors (PT-HWRs). One of the key stages in developing the thorium-based fuels for PT-HWRs is the reactor core configuration. Currently at CNL there are 20 core configurations under investigation, which involve several types of thorium-based fuels that could be implemented in a 700-MW(electric)-class PT-HWR. Among these core configurations, four fuel bundle concepts are being considered: (1) the reference (or nominal) 37-element bundle; (2) a 37-element modified bundle, with the center element using a different fuel material; (3) a 35-element bundle; and (4) an 18-element internally cooled annular fuel bundle. This study presents the steady-state subchannel thermal-hydraulic assessment of the 20 core configurations under investigation. The hottest channel approach is used in this study, as it represents the upper limit of a feasible design. The axial and element power distributions used in the analysis correspond to those of the discharge burnup. Three mass flows are considered in this study: 13.5, 21, and 24 kg/s. Five parameters are used to evaluate the fuel channel/bundle performance, namely, minimum critical heat flux ratio, channel pressure drop, enthalpy distribution, void fraction, and core power.

Neutronics and Safety Studies on a Research Reactor Concept for an Advanced Neutron Source

This paper presents preliminary neutronics and thermal hydraulics safety analysis results for a low-enriched uranium (LEU) fueled research reactor concept being studied at the National Institute of Standards and Technology (NIST). The main goal of this research reactor is to provide advanced sources for neutron scattering experiments with a particular emphasis given to high intensity cold neutron sources (CNSs). A tank-in-pool type reactor with an innovative horizontally split compact core was developed in order to maximize the yield of the thermal flux trap in the reflector area. The reactor concept considered a 20 MW thermal power and a 30-day operating cycle. For non-proliferation purposes, a LEU fuel (U3Si2-Al) with 19.75 wt% enrichment was used. The core performance characteristics of an equilibrium cycle with several representative burnup states—including startup and end of cycle—were obtained using the Monte Carlo–based code MCNP6. The estimated maximum perturbed thermal flux of the core is ~5.0 × 1014 n/cm2-s. The calculated brightness of the CNS demonstrates an average gain factor of ~4 compared to the current source operated at the existing NIST reactor. Sufficient reactivity control worth and shutdown margins were provided by hafnium control elements. Reactivity coefficients were evaluated to ensure negative feedback. Thermal hydraulics safety studies of the reactor were performed using the multi-channel safety analysis code PARET. Steady-state analysis shows that the peak cladding temperature and minimum critical heat flux ratio are less than design limits with sufficient safety margins. Detailed transient analyses for a couple of hypothetical design-basis accidents show that no fuel damage or cladding failure would occur with the protection of reactor scrams. All these study results suggest this new research reactor concept offers a demonstrable potential to greatly expand the cold neutron capability with a 20 MW power and certified LEU fuels.

Study of reactor plant disturbed cooling condition modes caused by the VVER reactor secondary circuit

Based on the equations of energy, mass and momentum conservation for the three-dimensional unsteady flow of the two-phase mixture, simulations were made of reactor plant disturbed cooling condition modes caused by the VVER secondary circuit using computer thermohydraulic codes. The use of computer codes to investigate these modes makes it possible to analyze transient and some emergency processes, without involving the industrial testing method; this helps create a basis for solving the problems of reliability, operational safety, and efficiency of nuclear power plants.A modern nuclear reactor is a complex system which cannot be investigated or calculated by using only simple theoretical models. Thermohydraulic calculations are an essential part of most of the design and technological developments in the nuclear power industry. Since the NPP conditions exclude the traditional technical way to check and specify the results and conclusions of an aprioristic analysis based on industrial testing, in some instances, it is possible to examine and predict the parameters of thermohydraulic processes in the reactor circulation circuit involving computer simulations.The primary objective of this work is to calculate and investigate reactor disturbed plant cooling condition modes caused by the VVER secondary circuit in order to determine whether the calculated parameters conform to the acceptance criteria established in the regulatory documents.Based on the RELAP-5, TRAC, and TRACE software codes, reactor plant cooling condition malfunction modes caused by the VVER-1000 secondary circuit were simulated and investigated. Experimental data on the mode with the turbine-generator stop valve closing are presented. The obtained dependences made it possible to determine the maximum values of pressure and temperature in the circulation circuit as well as estimate the Minimum Critical Heat Flux Ratio (MCHFR). It has been found that, if any of the initial events occurs, safety systems are activated according to the set points; transient processes are stabilized in time; and the Critical Heat Flux (CHF) limit is provided. Therefore, in the event of emergency associated with the considered modes, the reactor plant safety will be ensured.

Reactor Physics and Thermal Hydraulic Analysis of Annular Fuel Rod Cluster for Advanced Heavy Water Reactor

Abstract Internally and externally cooled annular fuel rod is an advantage over solid fuel rod in reactor core design in order to improve core power density and safety aspects like Minimum Critical Heat Flux Ratio (MCHFR) and lower centerline temperature. Various cluster geometries with annular fuel rod have been suggested for Advanced Heavy Water Reactor (AHWR) depending upon the operating parameters. We had already presented the design of 33 rod annular fuel cluster for AHWR in previous paper. The 33 pin fuel cluster had a lower MCHFR and the reactor power could be increased by about 20% only. The fuel bundle geometry presented in this paper is an improvised to obtain larger MCHFR and better physics performance. The AHWR uses a fuel cluster of 54 solid fuel rods and is designed for 920MWth. We have designed an annual fuel cluster with 19 fuel rods which meets all the thermal design margin as well as gives 30% more power than the solid fuel rod cluster without any considerable changes in design basis. Reactor core power calculations are carried out and then optimized burn up is obtained. This research article mainly deals with the improvements made in the annular fuel rod bundle to get optimized burn up and core power distribution.

Impact of Local Burnup on Prediction of Power Density in the NIST Research Reactor

This study addresses the overprediction of local power when the burnup distribution in each half-element of the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, is assumed to be uniform—a constraint in the full-core model used for neutronic analysis. A single-element model was utilized to quantify the impact of axial and platewise burnup on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and (proposed) low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, specifically, neutron energy spectrum, power distribution, and integral 235U vector. The power distribution changes significantly when incorporating local burnup effects and has lower power peaking relative to the uniform burnup case. In the uniform burnup case, the axial relative power peaking is overpredicted by as much as 59% in the HEU single element and 46% in the LEU single element. In the uniform burnup case, the platewise power peaking is overpredicted by as much as 23% in the HEU single element and 18% in the LEU single element. The degree of overprediction increases as a function of burnup cycle, with the greatest overprediction at the end of fuel element life. However, the overprediction in local power is always conservative in terms of the minimum critical heat flux ratio, a key safety parameter that depends on the local heat flux condition. The thermal flux peak is always in the midplane gap; this causes the local cumulative burnup near the midplane gap to be significantly higher than the fuel element average. Uniform burnup distribution throughout a half-element also causes a bias in fuel element reactivity worth particularly near end of life, primarily due to the importance of the fissile inventory in the midplane gap region. Despite this bias, comparisons of cycle length exhibit very good agreement between the core model with uniform burnup and the NBSR, which has many decades of operational experience with HEU fuel.

Design of 33 pin annular fuel rod cluster for Advanced Heavy Water Reactor

The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type reactor designed for utilizing thorium fuel with boiling light water as coolant and heavy water as moderator. The AHWR has several passive features, the major ones being heat removal through natural circulation, decay heat removal through passive means by the way of isolation condensers and a large heat sink in the form of gravity driven water pool. In the existing design of AHWR, the fuel rod cluster has solid fuel pins. The limitation on it is the Minimum Critical Heat Flux Ratio (MCHFR), which comes from coolant characteristics. In order to improve power output from the AHWR we are trying to design internally and externally cooled annular fuel rod cluster for AHWR. The fuel cluster design would have to be qualified from physics and thermal hydraulic behavior. In order to satisfy the design objectives, a detailed thermal hydraulics and physics simulation has been done.

Methodology for estimation of conditional probability of stagnation channel break for PHWRs

A methodology to estimate the conditional probability of a break in inlet feeder being the stagnation channel break in Pressurized Heavy Water Reactors is presented in this paper. According to AERB Safety Guide AERB/SG/D-5, rupture at any location of PHT system up to size of double ended largest feeder pipe belongs to category-2 events which are events of moderate frequency. For a particular range of break sizes in inlet feeder, the flow in the corresponding reactor channel reduces almost immediately to stagnant/very low flow conditions and remains sustained due to high flow inertia of PHWR pipe structure. As it is a single channel event out of hundreds of channels, the signal for reactor trip is generated very late. Consequently the fuel temperature rises sharply and fuel failure may occur. The probability of break in a feeder leading to flow stagnation conditions in the corresponding channel is low but needs to be estimated. In this paper the conditional probability of stagnation channel break for feeder pipe is estimated deterministically for a reactor similar to Indian PHWR. In this work, the stagnation channel break is identified if minimum critical heat flux ratio (MCHFR) is sustained lower than one anywhere in the corresponding reactor channel. It leads to sudden rise in fuel clad temperature due to sudden significant reduction in heat removal. There is a range of break sizes for which the clad surface temperature rises above fuel failure limits. Depending on this range and break locations in the feeder, the conditional probability is estimated. It is assumed that conditional probability depends on the range of break sizes leading to stagnation. It is observed from the calculations that the conditional probability of stagnation channel break is significant. The analysis has been carried out using thermal hydraulic code RELAP5/Mod 3.4.

Trip setpoint analysis for the reactor protection system of an advanced integral reactor

The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria.

Safety analysis for an inadvertent control rod withdrawal event during a power operation of an advanced integral reactor

To identify a safety margin in the case of an inadvertent control rod withdrawal event of a 65-MW t advanced integral reactor, safety analysis has been carried out by using the Transients And Setpoint Simulation/System integrated Modular Reactor (TASS/SMR) code. The diverse initial conditions, various reactivity insertion rates into a core, different combinations of a reactivity feedback and three different speed modes of a main coolant pump (MCP) have been considered to identify the effect of each parameter on a critical heat flux ratio (CHFR) and the initial condition resulting in the worst consequences from the viewpoint of the minimum critical heat flux ratio. The analysis results show that the worst consequences occur when a reactivity of 17.61 pcm/s is inserted into a core at an initial condition of a 45% initial core power, high coolant temperature at the core inlet position, low system pressure and a thermal design flow. It is also assumed that the least negative fuel and moderator temperature coefficients are applied. The safety parameters such as the minimum critical heat flux ratio and the system pressure are maintained within the safety limits and the reactor is safely transferred to a safe condition by a functioning of the safety systems of the advanced integral reactor.

A subchannel analysis of DUPIC fuel bundle for the CANDU reactor

Thermal characteristics of the reference DUPIC fuel has been studied for its feasibility of loading in the CANDU reactor. Half of the DUPIC fuel bundle has been modeled for a subchannel analysis of the ASSERT-IV Code which was developed by AECL. From the calculated mixture enthalpy, equilibrium quality and void fraction distributions in subchannels of the fuel bundle, it is found that the gravity effect may be pronounced in the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. The asymmetric distribution of the coolant in the fuel bundle is known to be undesirable since the minimum critical heat flux ratio can be reduced for a given value of the channel flow rate. On the other hand, the central region of the DUPIC fuel bundle has been found to be cooled more efficiently than that of the standard fuel bundle in the subcooled and the local boiling regimes due to the fuel geometry and the fuel element power changes. Based upon the subchannel modeling used in this study, the location of minimum critical heat flux ratio in the DUPIC fuel bundle turned out to be very similar to that of the standard fuel when the equivalent values of channel power and channel flow rate are used. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the subchannel-wise mixture enthalpy and void fraction peaks are located in the peripheral region of the DUPIC fuel bundle while those are located in the central region of the standard CANDU fuel bundle. Reduced values of the channel flow rates were used to study the effect of channel flow rate variation. The effect of the channel flow reduction on different thermal-hydraulic parameters have been discussed. This study shows that the subchannel analysis for the horizontal flow is very informative in developing new fuel for the CANDU reactor.

A numerical investigation of nuclear coupled density wave oscillations

A comprehensive mathematical model of boiling-water reactors (BWRs) for simulating nuclear coupled density wave oscillations is presented. A one-dimensional core dynamics program nars has been developed based on this model. A General Electric BWR/6 system is analyzed using nars. Our analysis shows that, even with very large power oscillations, the amplitude of heat flux oscillations is very low. The reduction in minimum critical heat flux ratio with large power oscillations is also observed to be quite low. Thus it can be concluded that the boiling transition does not occur even at very large power oscillations and hence fuel integrity is not compromised. It has been shown that void reactivity feedback has significant effect on the amplitude of the limit cycle oscillations and always has a destabilizing effect on the system. It is observed that the pressure loss coefficient in jet pumps has a significant effect on the reactor stability. Qualitative results obtained from a wide range of parametric studies on boiling channel stability together with the stability maps in terms of dimensionless numbers are presented. It has been observed that the power profile shape plays a very significant role in the stability of natural circulation loops.

Statistical evaluation of pin gaps in fuel clusters for fugen

The computer program STGAP has been developed to estimate pin gaps in a fuel assembly for FUGEN. The program optionally computes the probable distribution of the pin gap between any adjacent pair of fuel pins, either at a desired location in an assembly or longitudinally averaged over the total effective length of a pin, based on the measured manufacturing and assembling tolerances in geometrical dimensions and mechanical properties of all the independent elements composing a fuel assembly. It also correlates the computed fuel gap distribution with the minimum critical heat flux ratio in the corresponding local subchannel. Sample calculations were performed for the probable distributions of the pin gaps between pairs of adjacent fuel pins in the outermost layer of a FUGEN fuel assembly using the program and satisfactory agreement was obtained with the corresponding measured distributions.