Reactor Safety Research Articles

Transient characteristics analysis of residual heat removal system for Helium-Xenon mixture cooled small reactor system

Helium-Xenon (He-Xe) gas mixtures are characterized by stable physical properties, easy compression, and excellent thermodynamic performance, making them suitable for thermal cycle coolant for underwater reactors. Due to the unattended deep sea, and harsh and complex operating environment, the residual heat removal characteristics are directly related to reactors’ safety characteristics and stable and reliable operation ability. Therefore, studying the He-Xe gas mixture residual heat removal characteristics can provide an essential guarantee for reactor safety analysis. This article optimizes the physical property calculation model of the He-Xe gas mixture in the Brayton Cycle Reactor System Analysis Program (BRESA) and obtains a system property analysis program suitable for calculating the natural circulation of the He-Xe gas mixture. And a natural circulation system for the He-Xe mixed gas cooling reactor was constructed using the modified BRESA. Numerical calculations and theoretical analysis were conducted on the He-Xe gas mixture’s natural circulation flow process and residual heat removal characteristics. The study of the transient factors of temperature, flow rate, and other related parameters in the system circuit found that when the reactor is shut down, relying on the natural circulation system can effectively output the residual heat of the reactor core to ensure reactor safety. The relevant research results have important theoretical and scientific significance for the safety assessment and evaluation of the Brayton cycle reactor system.

Characterization of thermal stratification using validated liquid metal pool mixing models

Weakly stratified thermal fronts in the upper plena of liquid metal cooled reactors can pose prediction complexities for safety analyses. The strong thermal diffusivity of the coolant changes gravity’s influence on the flow by reducing horizontal density gradients at time scales faster than gravity can act on them. While damped out thermal behavior results in stratification, the underlying convection currents continue contributing to enhanced mixing. Experimental turbulence spectral information, captured by distributed velocity (ultrasonic) and temperature (fiber-optic) measurements, is used to construct microscale parameters and quantify the amount of enhanced vertical thermal diffusivity in a scaled gallium surrogate of a pool type sodium fast reactor upper plenum. These constructed microscale parameters are used to validate empirical models for use in liquid metal pools. This experimental data and, separately, the validated models, are fed into a simple 1-D transport model to quantify the impacts to reactor safety and allow for testing of the efficacy of using the more easily available bulk parameters (i.e., core barrel Reynolds and Richardson numbers) to estimate the enhanced diffusivity. The 1-D model represents the experimental fronts’ progressions most closely after accounting for these characteristics of mixed convection unique to liquid metals.

Neutronic and Thermal hydraulic evaluation of accident tolerant cladding materials in a WWER1000

The concept of Accident Tolerant Fuel (ATF) claddings is developed to improve the safety of light water reactors (LWRs). The main objective of this work is neutronic and Thermal Hydraulic (TH) analysis of potential cladding materials in a WWER1000. In this analysis, the ZrC, TiC, SiC and FeCrAl are evaluated as enhanced alternatives for traditional Zircaloy-4. For neutronic analysis, Monte Carlo particle transport method is used. For thermal hydraulic analysis, the performance of the materials in unmitigated small break LOCA is evaluated by MELCOR. Assessments show that ATFs can delay the failure of reactor pressure vessel (RPV) and core melting by reducing the heat generated from oxidation reaction. The TiC and FeCrAl claddings are more reliable in the accident, but they are not desirable from the perspective of the Neutronics. The calculations demonstrate that ZrC and SiC can be excellent alternatives to the Zircaloy-4. Also, simulations show using TiC and FeCrAl as a 0.1 mm thick covering of the Zircaloy-4 improves the multiplication factor, but it is not sufficient.

Integral and Separate Effects Test Facilities To Support Water Cooled Small Modular Reactors: A Review

This study reviews previous experimental facilities and test programs relevant to water-cooled reactor system design and analysis to meet regulatory compliances. This study aims to find the best solution for designing the required experiments, obtaining necessary test data, and verifying the developed computer code/models to support the new reactor design and development while minimizing cost and time while leveraging experiences from previous facilities to minimize. Nuclear reactor licensing requires supportive design, analysis, and experimental results to ensure the safety of the full-scale prototype reactor in regular operation, as well as during postulated accident scenarios. These reactor design analyses are generally performed using system codes and other associated simulation tools that require assessment, verification and validation using an appropriate experimental dataset. Experimental facilities used for reactor system safety analysis and system code assessments are categorized into integral effects test (IET) and separate effects test (SET) facilities. These IET and SET experiments and studies use geometrically scaled systems to reproduce the prototype system behavior at a reasonable cost, albeit with some scaling-related distortions. The design challenge of these model facilities is to identify and minimize scaling distortions while reproducing the most important operational phenomena in steady-state operation and in postulated accident scenarios. Lessons learned from previous experimental facilities, models, and correlations can support the development of new multipurpose, scaled, hybrid, integrated, and modular experimental facilities for advanced light water-cooled small modular reactors (SMRs). Successful operation of these facilities can significantly reduce upfront reactor development and demonstration costs and time to deployment.

CFD Uncertainty Quantification using stochastic spectral methods—Exemplary application to a buoyancy-driven mixing process

The consideration of uncertainties is of particular importance for nuclear reactor safety, where high safety standards for example ensure the integrity of the containment. By means of Computational Fluid Dynamics (CFD), buoyancy-induced mixing processes, which can take place during a severe reactor accident, are investigated. However, the CFD models contain uncertainties, which have a large impact on the present flow and have to be analyzed. The method development for the subsequent Uncertainty Quantification of a representative reactor test containment is conducted using the Differentially Heated Cavity of aspect ratio 4 with Ra=2×109 as a generic test case from the literature. In this way, methods for the future quantification of uncertainties in large-scale industrial applications are established. Results from single-fidelity models such as Unsteady Reynolds-Averaged Navier–Stokes, Large Eddy Simulation and Direct Numerical Simulation are presented and combined to three-level multifidelity models. Stochastic representations of scalar random responses are constructed by means of Polynomial Chaos Expansions and Karhunen–Loève Expansions are derived for the representation of stochastic processes. A new approach for the description of highly dynamic transient processes called Random Field Composition (RFC) is presented, which proposes stochastic model construction through combination of multiple random fields. The presented multifidelity (MF) models achieve high accuracy in combination with a justifiable computational effort. Therefore, MF modeling serves as a promising approach for large-scale applications. Furthermore, it is found, that RFC allows for the description of highly dynamic processes with a reasonable number of simulation runs, if the complexity of the stochastic process representation can be reduced by partitioning the stochasticity into single random fields.

Uncertainty quantification and target accuracy assessment of nuclear data to effective neutron multiplication factor of heat pipe cooled reactor

With the increasing improvement of the core physics calculation method, the uncertainty introduced by the calculation software and model approximation in the process of nuclear reactor core physics design is becoming smaller and smaller, while the uncertainty of the calculation results caused by the uncertainty of the input parameters itself is becoming more and more important. As an important input parameter of core physics calculation, the uncertainty of nuclear data will be transmitted, which will affect the accuracy of core physics calculation results. Quantifying the uncertainty of the core physics calculation results caused by nuclear data is an important element in the nuclear reactor safety design process. On the other hand, according to the target accuracy requirements of the nuclear reactor design, the key nuclear data and the priority of the improvement when the physical parameters of the nuclear reactor core reach the target accuracy requirements are evaluated, which can provide reference standards for the improvement of nuclear data and guide the development of specific experimental measurements work.

Preliminary neutronics and thermal analysis of a heat pipe cooled traveling wave reactor

Heat pipe cooled traveling wave reactor (HPTWR) is a novel concept of heat pipe reactor, which adopts neutron breeding wave to pursue Uranium-Plutonium breeding. In this paper, the cladding material, power output, and axial breeding fuel region length of the HPTWR are optimized in order to achieve the safe operation of reactor with reactor neutronics and HP thermal analysis. Monte Carlo neutron transport code RMC is used to obtain the neutronics results (reactivity, consumption and breeding of nuclear fuel, and power distribution). A heat pipe (HP) code based on thermal resistance–capacitance network model (TRCNM) is developed to analyze the heat transfer characteristics of HP using the power distribution of fuel element obtained by RMC code in the optimized HPTWR. Results show the optimized HPTWR can achieve 70 MWth power output and 46 years continuous operation. The highest temperature peak of HP wall in the highest power fuel element of reactor appears at beginning of cycle (BOC) and is about 1800 K. Due to the flattening of axial power distribution with propagation of traveling wave, the temperature peak of HP wall gradually decreases to about 1784.6 K at end of cycle (EOC), which can reduce the thermal stress of HP wall and improve the safety of reactor. The corresponding heat transfer margin of HP in the highest power fuel element is always above 19.0% during the whole reactor operation, which helps to ensure the safe operation of the optimized HPTWR.

Investigating the Potential of Nuclear Energy in Achieving a Carbon-Free Energy Future

This scientific paper discusses the importance of reducing greenhouse gas emissions to mitigate the effects of climate change. The proposed strategy is to reach net-zero emissions by transitioning to electric systems powered by low-carbon sources such as wind, solar, hydroelectric power, and nuclear energy. However, the paper also highlights the challenges of this transition, including high costs and lack of infrastructure. The paper emphasizes the need for continued research and investment in renewable energy technology and infrastructure to overcome these challenges and achieve a sustainable energy system. Additionally, the use of nuclear energy raises concerns, such as nuclear waste and proliferation, and should be considered with its benefits and drawbacks. The study assesses the feasibility of nuclear energy development in Latvia, a country in Northern Europe, and finds that Latvia is a suitable location for nuclear power facilities due to potential energy independence, low-carbon energy production, reliability, and economic benefits. The study also discusses methods of calculating electricity generation and consumption, such as measuring MWh produced by power plants, and balancing supply and demand within the country. Furthermore, the study assesses the safety of nuclear reactors, generated waste, and options for nuclear waste recycling. The transition to a carbon-free energy system is ongoing and complex, requiring multiple strategies to accelerate the transition. While the paper proposes that nuclear energy could be a practical means of supporting and backing up electricity generated by renewables, it should be noted that there are still challenges to be addressed. Some of the results presented in the paper are still based on studies, and the post-treatment of waste needs to be further clarified.

Key nuclear data for non-LWR reactivity analysis

An assessment of nuclear data performance for non-light-water reactor (non-LWR) reactivity calculations was performed at Oak Ridge National Laboratory that involved a thorough literature review to collect related observations made across different research institutions, an interrogation of the latest ENDF/B evaluated nuclear data libraries, and propagation of nuclear data uncertainties to key figures of merit associated with reactor safety for six non-LWR benchmarks. The outcome of this comprehensive study was published in a technical report issued by the US Nuclear Regulatory Commission. This paper provides a summary of the study’s key observations and conclusions and demonstrates with two examples how the various methods available in the SCALE code system were used to identify key cross section uncertainties for non-LWR reactivity analyses.

METHOD OF QUALIFICATION OF PASSIVE SAFETY SYSTEMS OF MODULAR NUCLEAR REACTORS WITH CIRCULATION CIRCUIT FLOWS

Low-power modular reactors are a promising direction for increasing the safety of nuclear power, because accident management in modular reactors is carried out only by passive safety systems (without electric pumps). Critical for the safety of modular reactors are accidents with a violation of the tightness of the natural circulation circuits of passive safety systems. The main limitations of using traditional accident modeling approaches with deterministic codes to qualify the reliability and operability of passive safety systems of modular reactors are related to the possibility of negative effects of "code differences" and "code user differences", as well as the unfoundedness of code verification/validation results. An original qualification method has been developed to ensure the safety conditions of the passive safety systems of the Westinghouse low-power modular reactor (SMR) in the event of accidents with a violation of the tightness of the natural circulation circuits. The assumptions adopted in the developed method ensure the conservatism of qualification results. Based on the preliminary calculation qualification of the natural circulation circuits of the SMR passive safety systems, it was established that for the relative sizes of the leaks, greater than 5% of the pipeline cross-section, a violation of safety conditions and drainage of the active zone may occur less than 24 hours after the start of the accident. It is necessary to modernize the SMR with regard to systems for diagnosing leaks in natural circulation circuits of passive safety systems and isolating damaged sections of circuits.

Verification of neutronics and thermal-hydraulic coupled system with pin-by-pin calculation for PWR core

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO2. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

Spray Cooling of Vertical Inline Horizontal Tubes

In water-cooled nuclear reactor safety, quench cooling is essential for minimizing core damage in the event of a Loss of Coolant Accident (LOCA). A hot, dry surface is quenched by bringing it into abrupt contact with a coolant, hence causing a fast fall in surface temperature. In this study, an air-atomized water spray cooling facility was built, and a series of tests were carried out to explore the quenching and rewetting of vertical inline horizontal tubes. During the cooling process, constant heat flux was provided on each tube. In the absence of a coolant, the steady heat flux represents the decay power of the channel components that reach the calandria tube surface. Temperature variations in the circumferential direction were measured over time during cooling. The effect of water pressure and air pressure on the rapid cooling of vertical inline horizontal tubes has been investigated. As air pressure rises, the quenching temperature of each tube decreases. A high-speed video camera was employed to investigate the flow behavior on the top tube surface as well as the spread of the re-wetting front around it. Experiments and visual analysis show that wetting front velocity rises with the water flow rate.

Numerical investigation on ballooning and rupture of a Zircaloy tube subjected to high internal pressure and film boiling conditions

Film boiling may lead to burnout of the heating element. Even though burnout does not occur, the heating element is subject to deformation because it is not sufficiently strong to withstand external loads. In particular, the ballooning and rupture of a tube under film boiling are important phenomena in the field of nuclear reactor safety. If the tube-type cladding of nuclear fuel ruptures owing to high internal pressure and thermal load, radioactive materials inside the cladding are released to the coolant. Therefore, predicting the ballooning and rupture is important. This study presents numerical simulations to predict the ballooning behavior and rupture time of a horizontal tube at high internal pressure under saturated film boiling. To do so, a multi-step coupled simulation of conjugated film boiling heat transfer and ballooning using creep model is adopted. The numerical methods and models are validated against experimental values. Two different nonuniform heat flux distributions and four different internal pressures are considered. The three-step simulation is enough to obtain a convergent result. However, the single-step simulation also successfully predicts the rupture time. This is because the film boiling heat transfer characteristics are slightly affected by the tube geometry related to creep ballooning.

The reliability analysis of pressurized water reactor safety injection system based on GO-FLOW

The safety injection system (RIS)is a large and complex fluid system in a nuclear power plant. Its central role is to ensure the safety of the core in the event of a breach event in the first circuit of the nuclear power plant. In order to investigate the timing of the safe injection system at the time of a Small-Break Loss-Of-Coolant Accident (SBLOCA), this paper uses the GO-FLOW method to model seven time points in the three operational phases of the safe injection system. The calculation is completed in conjunction with the failure rate of each component to give the reliability of the RIS at each point in time at SBLOCA. The final analysis yielded low reliability at the beginning of the direct cold section injection phase, which is of great reference for subsequent equipment maintenance and the safe operation of nuclear power.

Topography changes and microstructural evolution of nuclear graphite (IG-110) induced by Xe26+ irradiation

AbstractThe microstructure of nuclear graphite, a key material in nuclear reactors, is affected by the high-flux irradiation. The damage to the graphite by irradiation is important for reactor safety. To understand the damage of nuclear graphite by irradiation, IG-110 nuclear graphite, a representative nuclear graphite, was chosen to investigate the change in morphology and microstructure caused by 7 MeV Xe26+ irradiation with peak damage doses of 0.1-5.0 displacements per atom (dpa) for samples of a size of 40.0 mm×40.0 mm×2.0 mm. The topography and microstructure of IG-110 were characterized by SEM, AFM, grazing incidence XRD, Raman spectroscopy and nano-indentation. Results indicate that after 7 MeV Xe26+ irradiation at a dose of 0. 11 dpa, a ridge-like structure appears on the surface of the IG-110 graphite, mainly in the binder region, and the surface roughness increases slightly. With a further increase of the irradiation dose, the ridge-like structure also appears in the filler region. At a dose of 0.55 dpa, pore shrinkage increases accompanied by pore closure, and the surface roughness also increases. The changes in topography and microstructure caused by irradiation are attributed to the expansion of graphite along the c-axis direction. With increasing irradiation dose the defect density and the degree of in-plane disorder in the graphene sheets increases, while the modulus and hardness of the graphite first increase and then decrease. Their increase is caused by dislocation pinning and closure of fine pores, while their decrease is attributed to an increase in porosity and the generation of an amorphous structure.

DETERMINING GAMMA SOURCE IN URANIUM MOLYBDENUM OF FUEL IN G.A SIWABESSY MULTI PURPOSE REACTOR

Nuclear fission reactions produce a lot of radionuclides that release energy, one of which is in the form of gamma radiation. Gamma radiation is produced by various types of radionuclides, and nuclear reactor fuel will produce different values of gamma intensity. Uranium Molybdenum (U7Mo-Al) is the type of nuclear fuel for future research reactors that possesses many advantages. For the application of molybdenum-based fuel, it is necessary to determine the resulting gamma radiation. The purpose is to determine the gamma radiation produced from molybdenum-based fuel with various densities. This study begins with the determination of the mass composition of the reactor component, calculations with ORIGEN2.1, and data output analysis. The U7Mo-Al density was varied, namely 2.96 gU/cm3, 3.85 gU/cm3, 4.44 gU/cm3, 5.43 gU/cm3, 6.91 gU/cm3, and 8.29 gU/cm3. The gamma radiation yield of U7Mo-Al is lower than that of uranium silicide (U3Si2) with the same density of 2.96 gU/cm3. The result will add to the justification for the superiority of U7Mo-Al compared to U3Si2/Al. For U7Mo-Al with densities of 3.85 gU/cm3, 4.44 gU/cm3, 5.43 gU/cm3, 6.91 gU/cm3, and 8.29 gU/cm3, the one that produced the lowest gamma radiation intensity is 3.85 gU/cm3 while the highest is 8.29 gU/cm3. This explains that the intensity of the gamma radiation produced is directly proportional to the fuel density. The low intensity of gamma radiation in molybdenum-based fuel can be used as a suggestion in shielding design to ensure the operational safety of reactors.

Nuclear data uncertainty influence on the breeding ratio in sodium-cooled fast reactor systems

Nuclear data are a main uncertainty source in neutron transport simulations making their consideration in reactor safety necessary. This arises anew with the state-of-the-art reactors known as Generation IV. Some of the reactors suggests providing the reactivity margin below the effective delayed neutron fraction excluding prompt criticality accidents, and the breeding ratio is the key factor in this. Consequently, assessing a degree of the breeding ratio accuracy is of interest. Therefore, in this work, the breeding ratio uncertainties are analyzed by performing a sensitivity and uncertainty analysis of the MET1000 and MOX3600 models with respect to nuclear data using SCALE. As a result, the breeding ratio uncertainties are obtained approximately equal to 2% as the main contributors are 239Pun,γ, 238Un,γ, and 238Un,n'. The uncertainty sources between the models are compared, and 16O preponderantly increases the total uncertainty not directly by its uncertainty but by its impact on the spectrum.

Experimental Analysis of Steam Generator Tube Rupture in CIRCE Facility for MYRRHA Configuration

One of the main safety issues of Generation IV (Gen IV) heavy liquid-metal fast reactors is the postulated steam generator tube rupture (SGTR) accident. This event is characterized by primary and secondary coolant interaction, referred to in the literature as a coolant-coolant interaction event having a nonzero probability to occur. This accident scenario could affect the safety of a pool-type reactor, as a consequence of water secondary coolant flashing into the primary coolant liquid metal. The SGTR event needs to be experimentally characterized to evaluate the pressure waves effect, tube rupture propagation (domino effect), oxide precipitation and slug and plug formation, cover gas pressurization of the reactor, and steam flow paths through the pool and eventually the core, entailing the risk of positive reactivity insertion (due to positive local void coefficient). The design phase of the Gen IV MYRRHA plant has dealt with postulated SGTR safety issues in the framework of the MAXSIMA project, which is supported by the European Commission. A relevant contribution to this research activity was provided by the Italian Agency for New Technologies, Energy and Sustainable Economic Development Research Center Brasimone, where a new test section has been designed, assembled, instrumented, and implemented in the large-scale integral-effects pool facility CIRCE for investigating the SGTR event in a relevant configuration for the heat removal system of MYRRHA. This research reactor is not oriented to steam production for running a turbogenerator (no electric production), thus the heat removal system is referred to as the primary heat exchanger (PHX) and not as a steam generator. Four full-scale portions (four bundles of 31 tubes) of the MYRRHA PHX were adopted to carry out four independent SGTR experiments. Water flowed upward in the central tube of the bundle and two rupture positions were investigated at the lower and upper levels, named the bottom and middle scenarios, respectively. After the rupture, water was injected at 16 bar and 200°C into lead bismuth eutectic alloy at 350°C. The experimental results showed a remarkable repeatability and were presented in terms of (1) CIRCE vessel pressurization up to 2.7 and 4 bar absolute for the middle and bottom scenarios, respectively; (2) vapor flow paths through the bundle and its cooling effect up to 120°C and 140°C for the middle and bottom tests, respectively; and (3) strain measurements on tubes and bundle shells up to 2800 μm/m. The integrity of the tubes surrounding the ruptured one and the effectiveness of implemented safety device (rupture disks) pressure relief were significant engineering feedback for MYRRHA designers. The acquired high-quality data also constitute a database increase for future code verification and validation and possible new model development. The performed experimental analysis provided the awareness that a suitable design of a depressurization system (e.g., rupture disks) could allow for addressing postulated SGTR events in the MYRRHA configuration with confidence and safety.

Effects of base salt additives on NaCl–UCl3–PuCl3 fuel systems: Insights from CALPHAD simulations

The overall performance of molten salt reactor (MSR) fuels is governed by their chemical composition: the actinide content regulates fuel criticality; the base salt determines the fuel melting temperature, and the solubility of chemical species that evolves during the reactor’s lifetime. In this study, Calculation of Phase Diagrams (CALPHAD) simulations were performed to investigate the effects of non-unary base salts on the performance of NaCl–UCl3–PuCl3 fuel systems. We specifically focused on the fuel liquidus temperature and the fuel response to compositional changes, such as the breeding of PuCl3 and the evolution of fission products. Our calculations indicate that non-unary base salt combinations can effectively decrease the liquidus temperatures of the proposed fuel systems up to the eutectic actinide composition of the KCl–UCl3–PuCl3 system. Moreover, we found that the response of the fuel liquidus to shifts in fuel chemistry depends largely on the interactions between the base salt and actinide content. Our results imply that an appropriate modification of the base salt chemistry in the fuel can bolster the safety, performance, and economics of molten salt reactors.

Simulation of decay heat removal by active means following emergency shutdown in SFRs

Decay heat removal (DHR) is an important safety function in sodium-cooled fast reactors (SFRs). Dedicated systems are provided in the plant for performing DHR functionalities. Transients following reactor shutdown need to be studied to understand the dynamics of DHR systems, design the systems, and establish reactor safety. DHR systems can operate by active as well as passive means, thanks to the phenomenon of natural circulation. In the present work, a DHR system riding on the Balance of Plant (BoP) of a typical medium-sized SFR has been modeled and incorporated into the in-house developed plant dynamics code DYANA-P. DYANA-P already has models of primary and secondary circuits of SFR. With this code enhancement, the simulation of DHR in an integrated manner became possible. In particular, thermal models of the Moisture Separator Tank (MST), Decay Heat Removal Condenser (DHRC), and hydraulic model of the DHRC flow circuit were developed. Two-phase one-dimensional mass, momentum, and energy conservation equations with homogeneous flow assumptions were used to create the models. Simulation of SCRAM in a typical SFR has been carried out to demonstrate the applications of the modified code. Two different DHR deployment schemes were studied and compared – 1. Using DHR systems designed for high pressure of ∼170 bar, 2. Using DHR systems designed for low pressure of ∼15 bar. In the first scheme, reactor cooling was achieved at the desired rate of ∼20 K/h. In the second scheme, the reactor cooling rate will be higher at ∼60 K/h.