Decay Heat Removal Research Articles

Numerical analysis of forced convection heat transfer in a nuclear fuel storage vault

Forced convection heat transfer from fuel and blanket subassemblies to air in a nuclear fuel storage vault has been investigated in this work. Three dimensional simulations were performed for an actual fuel storage vault to improve the decay heat removal, air exchange rate, and thermal effectiveness. Forced airflow was used to remove the decay heat and an efficient ventilation arrangement was rigorously studied for long˗term storage of the nuclear fuel. The height of the interconnecting ducts between the blanket and fuel storage enclosures was systematically investigated to understand the ventilation performance. For this purpose, different temperature and air distributions inside the vault were analyzed. The mixing of air was enhanced, and the thermal stratification was reduced by connecting the ducts above the heat generating region. However, the fuel magazine temperature was reduced when the ducts were connected below the heat generating region. Furthermore, the effect of different shapes of interconnecting ducts on heat removal and air exchange rate through the vault was investigated. It was found that incorporating nozzles in the interconnecting ducts has led to an enhancement in air jet penetration and consequently a reduction in hot spot temperature. The validated computational model will be helpful in designing different aspects of nuclear fuel storage systems.

Turbulence model assessment and heat transfer phenomena inside a rectangular channel under forced and mixed convection

A very high-temperature gas-cooled reactor (VHTR) is an attractive option for hydrogen production due to its high operating temperature. The reactor cavity cooling system (RCCS), a passive safety system of the VHTR, surrounds the reactor vessel with rectangular riser channels and removes decay heat from it. Under emergency operation conditions such as a decrease in the flow rate of the RCCS, the flow condition inside the RCCS riser can become a mixed convection condition accompanying heat transfer deterioration. To design and operate the VHTR with reliable safety, accurate predictions for various convective heat transfer conditions in RCCS riser channels are required. In this study, by comparing the results of an air flow visualization experiment, the predictability of the turbulence models was assessed for the convective heat transfer and complex flow characteristics near the corner region of the rectangular duct confirmed from the experimental study. In this paper, the heat removal rate through the test section of the visualization experiment was quantified under various convective heat transfer conditions, and local temperature measurement data from the experiment were presented to describe the heat transfer deterioration phenomena inside a rectangular riser in turbulent mixed convection. Linear and nonlinear eddy viscosity models were evaluated by comparing the heat transfer coefficient, local temperature, local flow structure including secondary flow, and turbulence quantities obtained from the experiment. Furthermore, improvements to turbulence models for enhancing the prediction of mixed convection heat transfer in rectangular riser ducts were proposed.

Multi-tube effect of external condensation with non-condensable gas in a tube bundle

The passive safety design concept is essential to remove decay heat under the station blackout. We conducted experimental and CFD-based studies on condensation in the presence of non-condensable gas to analyze the performance of a Passive Containment Cooling System. External condensation experiments were performed in 2 ∼ 4 bar, air mass fraction in 0.2 ∼ 0.7 with 1 ∼ 3 cylindrical tubes. CFD simulation was also performed to analyze the condensation phenomenon considering the effect of the tube bundle in terms of the ratio of Heat Exchanger area and vessel volume (A/V). We conducted the sensitivity study on the various tube bundle and vessel geometry by CFD and suggested bundle correction parameters as a function of the A/V. As a result, we evaluated the PCCS design with the horizontal tube bundle. It significantly reduces the required number of tubes compared to vertical tube bundle design because of the larger tube surface area and higher HTC by an inclination enhancement factor.

Investigation into steam generator tube life expectancy affected by SFR transients

One of the decay heat removal systems in sodium cooled fast reactors (SFR) is the operation grade decay heat removal system (OGDHRS). During deployment of OGDHRS, steam generator has to operate at off-normal operating conditions which may lead to serious concerns with respect to flow stability in steam generator and the associated thermal fatigue on the tubes. Towards understanding this, tests are performed on a 19-tube once through steam generator simulating two OGDHRS deployment schemes. The consequent fatigue damage assessment of steam generator tubes is carried out using Finite Element Method based commercial software and RCC MR French design code. The fatigue damage fraction is estimated to be∼0.06 suggesting the acceptability of both the schemes for future Indian SFR. The effects of dry-out point temperature oscillation amplitude and its frequency on tube life are also studied. It is found that the oscillation frequency doesn't have much influence on the stress/strain range. It is further found that an increase in oscillation amplitude leads to a near linear increase in the stress range.

Performance evaluation of DRACS system of molten salt reactors using a transient solidification model

A transient solidification model was developed and coupled to ASYST code to simulate the transient behavior of the direct reactor auxiliary cooling system (DRACS) of the fluoride-salt-cooled high-temperature reactor (FHR) during the loss of heat sink (LOHS) accident. The model was firstly validated by comparing the experimental data with the simulation results that deriving from available heat transfer and friction factor correlations. The simulation results are found to be well-fitted with the experimental data. The transient solidification model successfully predicted the occurrence of solidification and the complete blockage of the heat transfer pipes in DRACSs. The solidified layer is found to be beneficial for the decay heat removal at the early stage of solidification, however, a sudden clog is found at the end stage of the solidification. Based on the simulation results, an optimization attempt was made to extend the failure time of DRACS by reducing the length of heat transfer tubes. The model is applicable to all the reactors where solidification may occur, such as other molten salt reactors (MSRs) and lead-cooled fast reactors (LFRs).

Natural convection phenomena in a liquid metal pool due to relocated and heap of heat-generating core debris: Numerical study

In the present numerical study, authors have investigated the natural convection heat transfer and fluid flow characteristics in a pool due to the heat generating core debris, i.e. decay heat in a typical fast breeder reactor. Two significant aspects of decay heat removal from debris under core disruptive accidents have been analyzed. These key aspects are (i) relocation of heat-generating core debris at the different locations inside the main vessel and (ii) effect of debris heap on heat transfer from source to ambient. These cases are analyzed extensively on full-scale three-dimensional reactor model with all major components. Multiple tray-type core catcher has incorporated with passive pipes over one collection tray. The present numerical study provides a robust and practical feasible core debris retention arrangement after a severe accident in fast breeder reactors. Also, the study is focused on a feasible array of core collection trays that maintained heat-generating debris and heat shield trays under safe thermal designed limit (∼1200 K for debris and ∼ 923 K for trays). This eliminates the further advancement of accidents and leakage of radioactive material into environments. During our analysis, it has been found that single tray with passive pipes maintained the heat-generating debris and tray under safe thermal limit for a fraction of 50 % destroyed core relocated to lower plenum while remaining heat-generating debris is present in its original position. Multiple tray arrangements accommodate the whole core debris within safe thermal design limits. With the possibility of debris heap after its relocation to the lower plenum, it has been found that heap angle less than 3° does not affect the integrity of core catcher.

Design of a direct-cycle supercritical CO2 nuclear reactor with heavy water moderation

A new reactor concept is described that directly couples a supercritical CO2 (sCO2) power cycle with a CO2-cooled, heavy water moderated pressure tube core. This configuration attains the simplification and economic potential of past direct-cycle sCO2 concepts, while also providing safety and power density benefits by using the moderator as a heat sink for decay heat removal. A 200 MWe design is described that heavily leverages existing commercial nuclear technologies, including reactor and moderator systems from Canadian CANDU reactors and fuels and materials from UK Advanced Gas-cooled Reactors (AGRs). Descriptions are provided of the power cycle, nuclear island systems, reactor core, and safety systems, and the results of safety analyses are shown illustrating the ability of the design to withstand large-break loss of coolant accidents. The resulting design attains high efficiency while employing considerably fewer systems than current light water reactors and advanced reactor technologies, illustrating its economic promise. Prospects for the design are discussed, including the ability to demonstrate its technologies in a small (∼20 MWe) initial system, and avenues for further improvement of the design using advanced technologies.

Modeling and numerical investigation on the effects of filling ratio in a large separate heat pipe loop

Large separate heat pipe is a potential selection to establish the passive cooling system to remove decay heat from the Spent Fuel Pool. In this system, heat pipes are expected to work properly and effectively for a long time without human intervention. Due to the physical dimensions and structures, flow patterns, and fluid distribution in large-scale heat pipes differ from those in conventional ones. Based on the condensation flow patterns, a flow condensation model is presented. Then, a separate heat pipe model considering filling ratio effects is developed. Using the new model, the mean absolute errors of simulation are 5.5% (water), 5.1% (ammonia), and 4.3% (R134a). While in the conventional condensation model, these errors are 29.4%, 9.7%, and 10%. It is found that the influence of filling ratio varies with working fluids. For water heat pipe, the optimal filling ratio is about 17%. Higher or lower filling ratios lead to degradation of heat transfer capacity. For R134a or ammonia, the optimal filling ratio is a range, about 20% to 70%. The boundary of the range relates to both evaporator outlet vapor quality and heat pipe downcomer liquid column height.

Numerical analysis of alternative operating temperatures on a modular helium reactor under accident conditions

An MHR (Modular Helium Reactor) can cool down by itself under any accident conditions even if active shutdown cooling systems are inoperative. The purpose of the study is to develop a numerical model to analyze the decay heat removal capabilities of an MHR. This study presents a comparison of coolant inlet/outlet operating temperatures (°C) for an MHR; (1)350/1000, (2)490/1000, (3)590/1000, (4)350/950, (5)490/950, and (6)590/950. The first objective is to predict the peak fuel temperature for different inlet/outlet temperatures under a depressurized cooldown accident. The second objective is to determine the effect of flow bypass gap size for all cases. To investigate the performance of different operating conditions and the sensitivity to gap size between graphite blocks, thermal-hydraulic transient simulations are performed using a commercial CFD code, Ansys Fluent. Passive cooldown under a depressurized accident is solved numerically for 100 h. The results indicate that all peak fuel temperatures are below the TRISO design limits of 1600 °C.

CRITICALITY SAFETY ANALYSIS OF THE DRY CASK DESIGN WITH AIR GAPS FOR RDNK SPENT PEBBLE FUELS STORAGECriticality Safety Analysis of the Dry Cask Design with Air Gaps for RDNK Spent Pebble Fuels Storage

Reaktor Daya Non-Komersial (RDNK) with a 10 MW thermal power has been proposed as one of the technology options for the first nuclear power plant program in Indonesia. The reactor is a High Temperature Gas-Cooled Reactor-type with spherical fuel elements called pebbles. To support this program, it is necessary to prepare dry cask to safely store the spent pebble fuels that will be generated by the RDNK. The dry cask design has been proposed based on the Castor THTR/AVR but modified with air gaps to facilitate decay heat removal. The objective of this study is to evaluate criticality safety through keff value of the proposed dry cask design for the RDNK spent fuel. The keff values were calculated using MCNP5 program for the dry cask with 25, 50, 75, and 100% of canister capacity. The values were calculated for dry casks with and without air gaps in normal, submerged, tumbled, and both tumbled and submerged conditions. The results of calculated keff values for the dry cask with air gaps at 100% of canister capacity from the former to the latter conditions were 0.127, 0.539, 0.123, and 0.539, respectively. These keff values were smaller than the criticality threshold value of 0.95. Therefore, it can be concluded that the dry cask with air gaps design comply the criticality safety criteria in the aforementioned conditions.

Experimental study on thermally assisted sagging deflection and interaction of multiple coolant channels

A reactor shutdown would typically initiate with a trip signal following a postulated Loss of Coolant Accident (LOCA). The Emergency Core Cooling System (ECCS) assures fuel cooling by flooding the core and recirculating coolant inside the channels for long-term decay heat removal from the fuel. When the primary flow is interrupted, and simultaneously the ECCS is severely impaired or failed, fuel cooling tends to degrade, leading to the heat-up of fuel and coolant channels. Analysis indicates that further progression of such accidents in Pressurized Heavy Water Reactors (PHWRs) is primarily delayed due to the huge water mass in the calandria vessel around the reactor core. However, the horizontal orientation introduces several phenomenological complexities introducing complexity in formulating the Severe Accident Management Guidelines (SAMGs). For a postulated unmitigated low-pressure accident scenario, the accident progresses with the channels heating up, and creep sagging under thermal load and channel and fuel weight. The sagged channel contacts the channel below, transferring heat and its load. With further progress, it is postulated that the supporting channel would fail, and leading to a cascading core collapse.The present paper discusses the experimental studies carried out on multiple channel sagging and disassembly phenomena envisaged during the mid-phase of a severe accident. It is challenging to perform full-length experiments; hence, scaled-down channels were used in the experiments. The postulated accident condition is a significant break LOCA with loss of ECCS and loss of all heat sinks. The experimental setup consists of three channels placed vertically in a fixed lattice pitch forming three distinct channel rows. The paper compares two possible situations or test cases: 1) all three channels are heated, 2) the top two channels are heated, and the lower channel is unheated. The lower channel is an unheated (dummy) channel supporting the upper deforming channels. Finally, a multi-channel sagging deformation mode has been postulated based on the experimental observations. The contact propagation along the contacting surface is found to be a time-dependent event, indicating a channel's rate-dependent elongation. The middle CT is compressed between the upper and lower channels, compromising its structural support for middle PT.

Computational Fluid Dynamic Analysis of the ESFR Reactor Pit Cooling System in Case of Sodium Leakage

Abstract The decay heat removal system (DHRS) for the European sodium-cooled fast reactor (ESFR) concept consists of three cooling systems, which provide highly reliable, redundant, and diversified decay heat removal functions. Two of the systems provide a strong line of defense, whereas the third system provides long term-heat removal. This third DHR system, DHRS-3, involves in separating oil, and water-cooled loops integrated in the reactor pit serves the purpose of the safety vessel. It is expected that the proposed DHR concept enables a robust demonstration of the practical elimination of the prolonged loss of the decay heat removal function. For its confirmation, detailed numerical analysis is needed as a basis for further investigation. Supporting this approach, the current evaluation with means of computational fluid dynamic (CFD) provides a preliminary thermal analysis of the capability of the oil cooling system in the reactor to be used for residual heat removal in the pit in case of emergency. For the evaluation, different heat flux values are assumed at the vessel wall to examine the range of the resulting temperatures. The temperature of the main vessel wall should be below 800 °C. Furthermore, a sodium leakage at 500 °C into the reactor pit is assumed. The concrete structure is below 70 °C.

Analysis of ESFR Decay Heat Removal Systems in Protected Station Blackout

Abstract As part of the European sodium fast reactor (ESFR)-safety measures assessment and research tools (SMART) project, safety assessments are being conducted on the updated ESFR design. An important part of the study is the evaluation of the reactor's behavior within hypothetical accidental conditions to assess and ensure that the accident would not lead to unexpected and disastrous events. In this paper, the analyzed accidental scenario is the so-called protected station blackout (PSBO), where the offsite power is lost for the power plant, simulated by using the TRACE and simulator-sodium fast reactor (SIM-SFR) system codes. The assessment started from the simulation of the reactor behavior without the decay heat removal systems (DHRSs). Following this, calculations of multiple DHRS arrangements have been performed to evaluate the individual and combined efficiency of the systems. Where it was possible, the results from the two system codes have been compared to show the consistency of the separate calculations. Through this study, the design of the DHRSs proposed at the beginning of the project has been investigated, and certain recommendations have been made for further improvement of the DHRS systems performance.

Evaluation of a freeze-tolerant decay heat removal system redundancy for fluoride salt-cooled high-temperature reactors (FHR)

Typical FHRs use a direct reactor auxiliary cooling system (DRACS) for decay heat removal under a variety of accident scenarios where active normal shutdown cooling is unavailable. Response surface method has been used for evaluating the reliability of the passive system. In this paper, a loss of heat sink (LOHS) accident sequence is evaluated using the response surface methodology. Response surfaces on maximum core outlet temperature and DRACS freezing time has been presented for evaluating the effects of friction loss, heat transfer coefficients and partial and full blockages of the DRACS. The present study shows that for one DRACS failure situations, the system is unconditional safe. But, for two DRACSs failure situations, there is probability that peak core outlet temperature (PCOT) will not exceed the allowable temperature. During the LOHS accident, the DRACS would not freeze within 3.33 days. This study will provide an insight for investigating DRACS redundancy and functional failure.

Three-dimensional numerical research on natural circulation characteristics of decay heat removal system in pool-type fast reactor CEFR

The pool-type sodium-cooled fast reactor decay heat removal system (DHRS) is an important engineered safety system to ensure the establishment of natural circulation in the case of station blackout (SBO). It is of great significance to study characteristics of DHRS for evaluating its passive residual heat removal capacity. Due to numerous core subassemblies and small scale of gaps between subassemblies, it is difficult to use the same scale grids to describe the three-dimensional flow paths in sodium pool with full core. Therefore few researches, either experiments or numerical simulation, have been performed on such complicated phenomena so far. In order to identify the natural circulation flow paths of fast reactor, an integrated three-dimensional simulation model of China Experimental Fast Reactor (CEFR) primary circuit system with full core was established, including 712 subassemblies and small gaps between subassemblies. FLUENT was used to carry out three-dimensional numerical simulation. Through the transient simulation of 5000 s, the main flow paths including the intra-subassembly flow path, the inter-wrapper flow path and the Main Vessel Cooling System (MVCS) backflow path were identified. At 5000 s, the intra-subassembly flow rate, the inter-wrapper flow rate, and the MVCS backflow rate were 3.07 kg/s, 1.53 kg/s, and −1.94 kg/s respectively. Thesymmetrical thermo-hydraulic characteristics of two loops were obtained to analyze thermal stratification in sodium pool. The results provided a key three-dimensional numerical reference for the safety of CEFR.

A Framework based on Finite Mixture Models and Adaptive Kriging for Characterizing Non-Smooth and Multimodal Failure Regions in a Nuclear Passive Safety System

In the safety analyses of passive systems for nuclear energy applications, computationally demanding models can be substituted by fast-running surrogate models coupled with adaptive sampling techniques; for speeding up the exploration of the components and system state-space and the characterization of the conditions leading to failure (i.e., the system Critical failure Regions, CRs). However, in some cases of non-smoothness and multimodality of the state-space, the existing approaches do not suffice. In this paper, we propose a novel methodological framework, based on Finite Mixture Models (FMMs) and Adaptive Kriging (AK-MCS) for CRs characterization in case of non-smoothness and/or multimodality of the output. The framework contains three main steps: 1) dimensionality reduction through FMMs to tackle the output non-smoothness and multimodality, while focusing on its clusters defining the system failure; 2) adaptive training (AK-MCS) of the metamodel on the reduced space to mimic the time-demanding model and, finally, 3) use of the trained metamodel provide the output for new input combinations and retrieve information about the CRs.The framework is applied to the case study of a generic Passive Safety System (PSS) for Decay Heat Removal (DHR) designed for advanced Nuclear Power Plants (NPPs). The PSS operation is modelled through a time-demanding Thermal-Hydraulic (T-H) model and the pressure selected for characterizing the PSS response to accidental conditions shows a strong non-smooth and multimodal behavior. A comparison with an alternative approach of literature relying on the use of Support Vector Classifier (SVC) to cluster the output domain is presented to support the framework as a valid approach in challenging CRs characterization.

New Reactor Safety Measures for the European Sodium Fast Reactor—Part I: Conceptual Design

Abstract Following up the previous collaborative project on European sodium fast reactor (CP-ESFR project), the European sodium fast reactor—safety measures assessment and research tools (ESFR-SMART project) considers the safety objectives envisaged for generation-IV reactors, taking into account the lessons learned from the Fukushima accident, in order to increase the ESFR safety level. In accordance with these objectives, guidelines have been defined to drive the ESFR-SMART developments, mainly simplifying the design and using all the positive features of sodium fast reactors (SFR), such as low coolant pressure, efficiency of natural convection, possibility of decay heat removal (DHR) by atmospheric air, high thermal inertia, and long grace period before a human intervention is needed. In this paper, a set of new ambitious safety measures is introduced for further evaluation within the project. The proposed set aims at consistency with the main lines of safety evolutions since the Fukushima accident. The paper gives a first review of the new propositions to enhance the ESFR safety, leading to a simplified reactor, forgiving and including a lot of passivity. This first version is supported by the various project tasks in order to assess the relevance of the whole design in comparison to the final safety objectives.

New Reactor Safety Measures for the European Sodium Fast Reactor-Part II: Preliminary Assessment.

The European Sodium Fast Reactor Safety Measures Assessment and Research Tools (ESFR-SMART) project offers innovative options for a sodium fast reactor (SFR) to improve its safety. This paper explains the preliminary calculations made on the main options to roughly verify their feasibility. Design propositions and calculations are here provided of the following innovative options: elimination of the safety vessel, innovative decay heat removal systems (DHRS), core catcher, thermal pumps, and secondary loops. In conclusion, all these options seem able to fulfill the key points of new safety rules for Generation-IV reactors. A status of the research and development (R&D) effort necessary to validate these new options is also proposed.

Studies on the optimal mitigation strategy of extended SBO for the isolated multi-unit NPP site

The Nuclear Power Plant in the isolated Multi-Unit site eventually reaches the core damage if it undergoes the extended Station Blackout with the depletion of available resources. However, the coping time can be maximized by the optimal strategy with proper operator’s actions so that the operator can prepare further mitigation action to prevent core damage. The resources-based mitigation strategy framework was developed from the insights of the probabilistic model and deterministic model, which consider monitoring of plant conditions, available decay heat removal systems, available power sources, and water sources depending on the time. The maximum coping time is estimated for various cases with MARS-KS1.4 code for APR1400 and combined with a corresponding probabilistic model for each accident sequence based on the accident conditions and the resources with additional human actions.

PRELIMINARY ASSESSMENT OF ENGINEERED SAFETY FEATURES AGAINST STATION BLACKOUT IN SELECTED PWR MODELS

The 2011 Fukushima accident did not prevent countries to construct new nuclear power plants (NPPs) as part of the electricity generation system. Based on the IAEA database, there are a total of 44 units of PWR type NPPs whose constructions are started after 2011. To assess the technology of engineered safety features (ESFs) of the newly constructed PWRs, a study has been conducted as described in this paper, especially in facing the station blackout (SBO) event. It is expected from this study that there are a number of PWR models that can be considered to be constructed in Indonesia from the year of 2020. The scope of the study is PWRs with a limited capacity from 900 to 1100 MWe constructed and operated after 2011 and small-modular type of reactors (SMRs) with the status of at least under licensing. Based on the ESFs design assessment, the passive core decay heat removal has been applied in the most PWR models, which is typically using steam condensing inside heat exchanger within a water tank or by air cooling. From the selected PWR models, the CPR-1000, HPR-1000, AP-1000, and VVER-1000, 1200, 1300 series have the capability to remove the core decay heat passively. The most innovative passive RHR of AP-1000 and the longest passive RHR time period using air cooling in several VVER models are preferred. From the selected SMR designs, the NuScale design and RITM-200 possess more advantages compared to the ACP-100, CAREM-25, and SMART. NuScale represents the model with full-power natural circulation and RITM-200 with forced circulation. NuScale has the longest time period for passive RHR as claimed by the vendor, however the design is still under licensing process. The RITM-200 reactor has a combination of passive air and water-cooling of the heat exchanger and is already under construction.