Thermal-hydraulic modeling of flow inversion in a research reactor

The course of loss of flow accident and flow inversion in a pool type research reactor, with scram enabled under natural circulation condition is numerically investigated. The analyses were performed by a lumped parameters approach for the coupled kinetic–thermal-hydraulics, with continuous feedback due to coolant and fuel temperature effects. A modified Runge–Kutta method was adopted for a better solution to the set of stiff differential equations. Transient thermal-hydraulics during the process of flow inversion and establishment of natural circulation were considered for a 10-MW IAEA research reactor. Some important parameters such as the peak temperatures for the hot channel were obtained for both high-enriched and low enriched fuel. The model prediction is also verified through comparison with other computer code results reported in the literature for detailed simulations of loss of flow accidents (LOFA) and the agreement between the results for the peak clad temperatures and key parameters has been satisfactory. It was found that the flow inversion and subsequent establishment of natural circulation keep the peak cladding surface temperature below the saturation temperature to avoid the escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation to ensure the safe operation of the reactor.     

Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes.     

Sensitivity of reactivity insertion limits with respect to safety parameters in a typical MTR

The sensitivity of various safety parameters, affecting the reactivity insertion limits imposed by clad melting temperature for a typical pool type research reactor, have been investigated in this work. The analysis was done for low enriched uranium (LEU) core with scram disabled conditions. The temperature coefficients of fuel and coolant, void/density coefficient and β_eff were individually varied and the reactor behavior for different ramp reactivity transients was studied. In this work ramp reactivity insertions from 1.6 to 2 $/0.5 s were selected and peak power, maximum fuel, clad and coolant temperatures were determined. Results show that peak power decreases with an increase in the Doppler coefficient of reactivity. However, it rises with an increase in the reactivity insertion. Core remains insensitive to the coolant temperature coefficient of reactivity for ramps in the range of 1.6–1.9/0.5 s. Peak power decreases with an increase in the void coefficient of reactivity (0.1 $/%void to 0.8 $/%void). With a decrease in the void coefficient of reactivity, the maximum fuel and clad temperatures show a non-linear rise. Power and temperature peaks in the transient are sensitive to the values of β_eff. Finally, it can be concluded that LEU is a safe core due to its smaller β_eff, larger Doppler coefficient and void coefficient of reactivity. It is inferred through this work that reactivity insertion limits of LEU core are quite insensitive to β_eff, the Doppler coefficient and the coolant temperature coefficient of reactivity. They are highly sensitive to the change of the void coefficient of reactivity in the core.     

DHR-200池式低温供热堆SBO-ATWS事故及自然循环能力分析

DHR-200池式低温供热堆(简称DHR-200池式堆)设计有自然循环瓣阀,为检验其安全性,选取典型的全厂断电叠加紧急停堆系统失效(SBO-ATWS)事故,使用RELAP5程序对其热工水力参数瞬态特性及其自然循环能力进行分析。结果表明,DHR-200池式堆具有很好的负温度反应性反馈效应,即SBO-ATWS事故后,由于燃料和冷却剂温度升高,引入负反应性,可使反应堆实现热停堆;事故后,通过非能动方式开启自然循环瓣阀,可建立稳定的自然循环,将堆芯衰变热导出至堆水池内,验证了DHR-200池式堆的固有安全性。     

Analysis of SBO-ATWS Accident and Natural Circulation Capacity of DHR-200 Pool Type Low Temperature Heating Reactor

DHR-200 Pool Type Low Temperature Heating Reactor (DHR-200) was designed with natural circulation flap valve. In order to examine the safety of the DHR-200, the RELAP5 code was used to analyze the transient thermal-hydraulic characteristics and the natural circulation capacity under the station blackout anticipated transient without scram (SBO-ATWS). The results show that DHR-200 has enough negative temperature reactivity feedback effect. With the rising of the temperatures of the fuel and the coolant, finally the reactor can be shut down by the effect of the negative temperature reactivity feedback effect. After the accident, the natural circulation flap valve will be opened by passive means to establish a stable natural circulation, and then the residual heat of the core can be removed to pool of the reactor. Therefore, it is demonstrated that the DHR-200 has good inherent safety features.     

Prospects of using different clad materials in a material test research reactor – Part 3 – The dynamic behavior

The effects of using different clad materials on the dynamics of a material test research reactor were studied. For this purpose, the aluminum clad of an MTR was replaced separately with stainless steel-316 and zircaloy-4. Simulations were carried out to determine the reactor performance under reactivity insertion and loss-of-flow transients. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that during the fast reactivity insertion transient, Al cladded fuel attained the maximum reactor power of 59.34 MW, while stainless steel-316 cladded attained 48.74 MW and zircaloy-4 cladded attained maximum power of 55.87 MW. During the slow reactivity insertion transient, Al cladded fuel attained the maximum reactor power of 12.38 MW, while stainless steel-316 cladded attained 12.23 MW and zircaloy-4 cladded attained maximum power of 12.34 MW. During the loss-of-flow transients, the reactor power of the stainless steel-316 cladded fuel remained slightly lower than the other two. The fuel temperature of stainless steel-316 and zircaloy-4 cladded fuels remained higher due to poor fuel–clad gap conductance.     

Simulation of reactivity transients in current MTRs

This work aims at simulation of reactivity induced transients in High Enriched Uranium (HEU) and Low Enriched Uranium (LEU) cores of a typical Material Test research Reactor (MTR) using PARET code. The transient problem was forced through specification of externally inserted reactivity as a function of time. Reactivity insertions are idealized by ramps and steps. Superdelayed-critical transients, superprompt-critical transients and quasistatic transients are selected for the analysis. Ramp and step reactivity functions were employed to simulate these perturbations. The effect of initial power on transient behavior has also been investigated. The low enriched uranium core is analyzed for transients without scram. The magnitudes of maximum reactivity insertions are chosen to be in the range of $0.05 to 2.0 for different reactivity insertion times. Transient simulation with scram reveals that response of both HEU and LEU-cores is similar for selected ‘ramps’ and ‘steps’. The difference is observed in the peak values of power and coolant, clad and fuel temperatures. Trip level is achieved earlier in case of LEU-core. The peak clad temperatures in both LEU and HEU-cores remain below the melting point of aluminum-clad for the selected reactivity insertions. Simulation show that the LEU-core is more sensitive to perturbations at low power as compared to the transients at full power. For reactivity transients at low power level, power rises sharply to a higher peak value. In transients at full power, the peak power barely exceeds the trip level. The power oscillations after the first peak are observed for transients without scram.     

Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT

The IAEA’s reference research reactor MTR-10 MW has been modeled using the code MERSAT. The developed MERSAT model consists of detailed representation of primary and secondary loops including reactor pool, bypass, main pump, heat exchanger and reactor core with the corresponding neutronics and thermalhydraulic characteristics. Following the successful accomplishment of the steady state operation at nominal power of 10 MW, reactivity insertion accident (RIA) for three different initial reactivity values of $1.5/0.5 s, $1.35/0.5 s and $0.1/1.0 s have been simulated. The predicted peaks of reactor power, hot channel fuel, clad and coolant temperatures demonstrate inherent safety features of the reference MTR reactor. Only in case of the fast RIA of $1.5/0.5 s, where the peak power of 133.66 MW arrived 0.625 s after the start of the transient, the maximum hot channel clad temperature arrives at the condition of subcooled boiling with the subsequent void formation. However, due to the strong negative reactivity feedback effects of coolant and fuel temperatures the void formation persists for a very short time so that thermalhydraulic conditions remained far from exceeding the safety design limits of thermalhydraulic instability and DNB. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermalhydraulic system codes.     

Reactor thermal behaviors under kinetics parameters variations in fast reactivity insertion

The influences of variations in some of the kinetics parameters affecting the reactivity insertion are considered in this study, it has been accomplished in order to acquire knowledge about the role that kinetic parameters play in prompt critical transients from the safety point of view. The kinetics parameters variations are limited to the effective delayed neutron fraction (β_eff) and the prompt neutron generation time (Λ). The reactor thermal behaviors under the variations in effective delayed neutron fraction and prompt neutron generation time included, the reactor power, maximum fuel temperature, maximum clad temperature, maximum coolant temperature and the mass flux variations at the hot channel. The analysis is done for a typical swimming pool, plate type research reactor with low enriched uranium. The scram system is disabled during the accidents simulations. Calculations were done using PARET code. As a result of simulations, it is concluded that, the reactor (ETRR2) thermal behavior is considerably more sensitive to the variation in the effective delayed neutron fraction than to the variation in prompt neutron generation time and the fast reactivity insertion in both cases causes a flow expansion and contraction at the hot channel exit. The amplitude of the oscillated flow is a qualitatively increases with the decrease in both β_eff and Λ.     

Cobalt irradiation box ejection accident of ETRR-2

The new Egyptian Test and Research Reactor Number 2 ETRR-2, MTR type, is now under operational tests. It has a main central irradiation channel for the purpose of Co⁶⁰ isotope production with an intended rated capacity of 50 000 Ci per year. The reactivity introduced in the reactor due to accidental ejection of the Co⁶⁰ irradiation box (CIB) should be discussed. This reactivity insertion accident (RIA) may be fast or slow with maximum reactivity worth 2.9428 $. The CIB may move with constant speed or variable acceleration according to its initial speed and the applied forces. This results in a linear, parabolic or sinusoidal motion, which in turn affects the reactivity insertion rate (RIR). The present work analyzes this type of perturbation during normal operating conditions: 22 MW full power and 1900 kg s⁻¹ forced core cooling flow. The work serves as a part of the safety evaluation process applicable to similar MTR cores. The RIA code TRANSP20 is developed for this study. It simulates various types of RIR, fast or slow resulting from different CIB ejections. Scram signal due to power, period, inlet and outlet temperatures, or temperature difference is expected to activate the shutdown system. The work presents five case studies, two for fast ejection and three for slow. The transient behavior of the reactor during this is illustrated. The results show that the reactor can withstand slow ejection if the scram is available. However, for fast ejection the scram system does not prevent the clad temperature from exceeding safety limits. Recommendations to prevent or mitigate this accident are highlighted.     

Thermal-hydraulic behavior of the NUR nuclear research reactor during a fast loss of flow transient

Nuclear safety analysis remains of crucial importance for both the design and the operation of nuclear reactors. Safety analysis usually entails the simulation of several selected postulated accidents, which can be divided into two main categories, namely reactivity insertion accident (RIA) and loss of flow accident (LOFA). In this paper, thermal-hydraulic simulations of fast LOFA accident were carried out on the new core configuration of the material test research reactor NUR. For this purpose, the nuclear reactor analysis PARET code was used to determine the reactor performance by calculating the reactor power, the reactivity and the temperatures of different components (fuel, clad and coolant) as a function of time. It was observed that during the transient the maximum clad temperature remained well below the critical temperature limit of 110 °C, and the maximum coolant temperature did not exceed the onset of nucleate boiling point of 120 °C. It is concluded that the reactor can be operated at full power level with sufficient safety margins with regard to such kind of transients.     

Effects of high density dispersion fuel loading on the dynamics of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the dynamics of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different properties affecting the reactor in different ways, fuels U–Mo (9w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to determine the reactor performance under reactivity insertion and loss of flow transients. Nuclear reactor analysis code PARET was employed to carry out these calculations. It is observed that during the fast reactivity insertion transient, the maximum reactor power is achieved and the energy released till the power reaches its maximum increases by 45% and 18.5%, respectively, as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved during the transient, by 27.7 K, 19.7 K and 7.9 K, respectively. The time required to reach the peak power decreases. During the slow reactivity insertion transient, the maximum reactor power achieved increases slightly by 0.3% as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³ but the energy generated till the power reaches its maximum decreases by 5.7%. The temperatures of fuel, clad and coolant outlet remain almost the same for all types of fuels. During the loss of flow transients, no appreciable difference in the power and temperature profiles was observed and the graph plots overlapped each other.     

49-2泳池堆低温供热全厂断电ATWS事故分析

针对49-2泳池式反应堆（简称49-2泳池堆）用于城市低温供热的工况,选取典型的全厂断电叠加紧急停堆系统失效(全厂断电ATWS)的超设计基准事故,使用RELAP5/MOD3.2程序对其热工水力参数瞬态特性进行分析。结果显示,49-2泳池堆具有很好的负温度反馈效应,事故后,由于燃料和冷却剂温度升高,从而引入一定的负反应性,使反应堆处于次临界状态;同时堆芯通过与堆水池建立自然循环,将衰变热带出,最终依靠自然循环方式将堆芯余热排出至上部大气环境热阱,验证了49-2泳池堆用于城市低温供热的固有安全性。     

Analysis of reactivity induced accidents at Pakistan Research Reactor-1

Analysis of reactivity induced accidents in Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel, has been carried out using standard computer code PARET. The present core comprises of 29 standard and five control fuel elements. Various modes of reactivity insertions have been considered. The events studied include: start-up accident; accidental drop of a fuel element on the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results reveal that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is concluded that the reactor, which is operated safely at a steady-state power level of 10 MW, with coolant flow rate of 950 m³/h, will also be safe against any possible reactivity induced accident and will not result in a fuel failure.     

Post-test analysis of helium circulator trip without scram at 3 MW power level on the HTR-10

Safety demonstration tests were conducted on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the transient data of reactor core and primary cooling system for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 3 MW power level on October 15, 2003. This paper simulates and analyzes the power transient and the thermal response of the reactor during the test by using the THERMIX code. The analytical results are compared with the test data for validation of the code.Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shut down after the stop of the helium circulator; the subsequent phenomena such as the recriticality and power oscillations are also studied. During the test a natural circulation loop of helium is established in the core and the other coolant channels and its consequent thermal response such as the temperature redistribution is investigated. In addition, temperatures of the measuring points in the reactor internals are calculated and compared with the measured values. Satisfactory agreements obtained from the comparison demonstrate the basic applicability and reasonability of the THERMIX code for simulating and analyzing the helium circulator trip ATWS test. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature during the test is always lower than 1600 °C which is the limited value for the HTGR.     

Thermal hydraulic and safety analysis for core conversion (HEU–LEU) of Syrian Miniature Neutron Source Reactor

The paper presents the behavior and properties analysis of the low enriched uranium fuel compared with the original high enriched uranium fuel. The MNSR reactor core was modeled with both fuel materials and the reactor behavior was studied during the steady state and abnormal conditions. The MERSAT code was used in the analysis. The steady state thermal hydraulic analysis results were compared with that obtained from the experimental results hold during commissioning the Syrian MNSR. Comparison with experimental data shows that the steady-state behavior of the HEU core was accurately predicted by the MERSAT code calculations. The validated model was then used to analyze LEU cores with two proposed UO₂ fuel pin designs. With each LEU core, the steady state and 3.77 mk rod withdrawal transient were run and the results were compared with the available published data in the literatures for the low enriched uranium fuel core. The results reveal that the low enriched uranium fuel showed a good behavior and the peak clad temperatures remain well below the clad melting temperature during reactivity insertion accident.     

Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U–Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm³, 7.74 gU/cm³ and 8.57 gU/cm³. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.     

Thermal-hydraulic analyses of transients in an actinide-burner reactor cooled by forced convection of lead–bismuth

The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead–bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.     

Estimation of fuel and clad temperature of a research reactor during dry period of de-fuelling operation

Heat transfer and fluid flow studies related to spent fuel bundle of a research reactor in fuelling machine has been carried out. When the fuel is in reactor core, the heat generated in the fuel bundle is removed by heavy water under normal reactor operation. However, during the de-fuelling operation, the fuel bundle is exposed to air for some period called dry period. During this period, the decay heat from fuel bundle has to be removed by air flow. This flow of air is induced by natural convection only. In this period, the temperatures of fuel and clad rise. If clad temperature rises beyond a certain limit, structural failure may occur. This failure can result into release of fission products from fuel rod. Hence the temperature of clad has to be within specified limit under all conditions. The objective of this study is to estimate the clad temperature rise during the dry period.In the CFD simulation, the turbulent natural convection flow over fuel and radiation heat transfer are accounted. Standard k-? model for turbulence, Boussinesq approximation for computing the natural convection flow and IMMERSOL model for radiation are used.The steady state and transient CFD simulation of flow and heat is performed, using the CFD code PHOENICS. The steady state analysis provides the maximum temperature the clad will attain if fuel bundle is left exposed to air for sufficiently long time. For safe operation, the clad temperature should be limited to a specified value. From steady state CFD analysis, it is found that steady state clad temperature for various decay powers is higher than the limiting value. Hence transient analysis is also performed. In the transient analysis, the variation of clad temperature with time is predicted for various decay powers. Safe dry time, i.e. the time required for clad to reach the limiting value, is predicted for various decay powers. Determination of safe dry time helps in deciding the time available to the operator to drop the bundle in light water pool for storage. The analysis is found useful in optimizing the de-fuelling process.