Main Heat Transport Research Articles

Main heat transport system configuration influence on steam drum level control and safety for a pressure tube type boiling water reactor with multiple interconnected loops

Abstract NCABWR (Natural-Circulation-Advanced-Boiling-Water-Reactor) is a pressure-tube-type BWR with multiple inter-connected Main-Heat-Transport (MHT) system loops and steam drums (SDs). SD-Level-Control is related to, the MHT coolant inventory and sustained heat removal, hence the overall safety of the Nuclear-Power-Plant (NPP). SD-Level is vital for ensuring presence of adequate coolant-inventory with sufficient and prolonged cooling of fuel, for normal-operation and accidents. In the initial-MHT-configuration of NCABWR the four MHT-loops are, inter-connected at the Common-Reactor-Inlet-Header (CRIH), which influences the SDLC adversely; such occurrences are, not reported in the open literature. Open loop (without SD-level-controllers) step-change response showed that initial MHT-configuration response is stable, non-oscillatory and non-diverging even for a high-magnitude step-change in the feed-flows. Closed-loop (with SD-level-controllers) response, even to minor disturbances leads to diverging SD-Levels due to CRIH-MHT-configuration (Gaikwad, A.J., Vijayan, P., Iyer, K., Bhartiya, S., Kumar, R., Lele, H.G., Ghosh, A.K., Kushwaha, H.S., and Sinha, R.K. (2009). Effect of loop configuration on steam drum level control for a multiple drum interconnected loops pressure tube type boiling water reactor. IEEE Trans. Nucl. Sci. 56: 3712–3725). CRIH is required to overcome Postulated-Initiating-Event (PIE) for single-accumulator-failure-LOCA. A hypothetical MHT-configuration with partitioning of CRIH will isolate each loop from one another and lead to a stable SDLC, but then for the above-mentioned PIE safety is not met. LOCA in one of the four ECCS header-compartments require a mandatory interconnection among, the four MHT-Loops at CRIH. Solution to SDLC issue were explored with changes in MHT-configuration and with innovative solution-schemes for feed-water-system and SDLCs. In the present study the pioneering solution-schemes with individual SD pressure control for SD level control and the Split/Double baffle are proposed, implemented and the performance was tested with simulation are discussed in details with other solution schemes.

A random 2D walk of a submerged free-floating disc in a convective layer

Free motions of a thermally insulating disc submerged in a liquid layer heated from below are studied experimentally. The disc is fixed at a certain height from the bottom of the convective cell, but can move freely in horizontal directions. Blocking of vertical motion and heat flux leads to the appearance of large-scale vortices, which in turn, drags the disc. The interplay of the large-scale circulation and the disc produces complex two-dimensional disc motions. The disc dynamics is mainly determined by large-scale modes: toroidal vortex with descending fluid in the centre and large-scale vortices along each coordinate, existing on the background of intense, but relatively short-lived small-scale convective cells. The wandering disc forms a cold spot above itself, which affects large-scale circulation but has little effect on cell convection at scales on the order of the layer thickness, which provide the main heat transport and dominate the upper part of the layer. Despite the essentially three-dimensional flow structure, which is fundamentally different from that of the convective flow in an elongated tank with one-dimensional disc motions, the travels along each coordinate look similar to one-dimensional motions.

Ensuring safety of new, advanced small modular reactors for fundamental safety and with an optimal main heat transport systems configuration

Abstract Many countries are considering Small and Modular Reactors as a viable alternative to counter the climate-change/global-warming with a quick deployment of green, carbon free nuclear energy option in the energy mix. Proponents of SMRs claim that these designs rely more on enhanced inherent/engineered safety and passive features with novel concepts. SMRs are being designed to be fabricated at a factory and then transported as ‘modules’ to the sites for installation either as a single module or multiple module plant. There are many variant of SMRs under considerations/design/construction/commissioning/operation states and majority of the, more than 70 odd SMRs are in the design stage. The paper focuses on safety aspects while addressing the fundamental safety requirement that are derived from fundamental safety principles, the acceptance criteria, the expected/envisaged safety targets and not only the economic impact/considerations. The assessment basis for requirements towards safety enhancements and their extent of assurance in the design are highlighted against the claims made. Ensuring SMR safety with respect to the fundamental safety functions will depend on the foreseen/predicted fission product releases, following overheating of the fuel, during the worst/credible accident conditions and likelihood of occurrence of these accidents. Innovations in the development of advanced fuel, deploying passive safety systems, novel concepts in main heat transport system configuration and advanced features in instrumentation can help in realising the goal of ensured enhanced safety in the SMRs, both in preventive and mitigation domains during severe accidents. Enhancements in the acceptance criteria and deterministic and probabilistic safety targets is also expected and may be envisaged. The paper brings out the challenges faced in the design and regulation of the new NPPs, while addressing fundamental safety principles implementation, generic, specific safety issues, and only genuine innovations can ensure and improve the safety. Aspects related to passive systems and the optimal main heat removal system configuration of the NPPs are also discussed. The aspects related to concurrent design and regulation of new NPPs including SMRs also has been brought out in the paper.

Evaluation of elevated temperature process for the decontamination of stainless steel 304LN surface

Abstract Deposition of activated corrosion products on out-of-core surfaces and ensuing radiation field build-up in nuclear reactors calls for remedial measures like decontamination to minimize exposure of service personnel carrying out routine maintenance jobs. Decontamination of stainless steel structural materials in nuclear power plants pose challenges due to the inherent difficulty in dissolving chromium containing oxides like ferrites or chromites. Stainless steel, SS 304LN, is the proposed structural material for the main heat transport system of Indian Advanced Heavy Water Reactors (AHWRs) and it is important to be equipped with an effective decontamination process to mitigate the radiation field build-up in AHWRs. This paper discusses the studies to develop an elevated temperature (ET) decontamination process for SS304LN surface. Oxide covered SS304LN specimens were prepared under simulated AHWR coolant chemistry conditions. The oxide film formed on the specimen surface was Ni and Cr substituted single phase ferrite spinel. The oxide film could be completely dissolved from the SS304LN surface in the nitrilotriacetic acid (NTA) based ET process at 180 °C in a single step with an acceptable and low corrosion rate of 7.7 × 10−3 μm/h. The efficiency of the ET process was compared with the known NTA based conventional low temperature dilute chemical decontamination process and a multi-step, multi-cycle process comprising of alternate oxidation and reduction processes. The ET process proved to be more effective and apt for decontamination of SS304LN surface.

Optimal Main Heat Transport system configuration for a nuclear power plant

Selection of an optimal Main Heat Transport (MHT) system configuration for a nuclear power plant (NPP) is an activity of high significance, as it influences the safety and operational aspects. This paper brings out the methodology for selection of optimal MHT configuration of NPPs which provides enhanced safety along with ease of operation. The criterion for selection of NPP-MHT configuration have been developed for enhancing safety and ease of operation. As a case study, an innovative MHT configuration has been arrived for a channel type Boiling Water Reactor (BWR) which further improves safety. Some of the multiple failure Design Extension Conditions (DEC) scenarios without SCRAM which lead to severe core damage, for commonly deployed MHT configurations, do not encounter any core heat up with this optimal MHT configuration, for a prolonged duration with Emergency Core Cooling System (ECCS). To demonstrate the effect of MHT system configuration on safety, the Natural Circulation Advanced Boiling Water Reactor (NCABWR) has been considered in this study. Though NCABWR is considered for demonstration purpose, similar methodology can be used for devising or conceptualizing an optimal MHT system configuration for any other NPP. The NCABWR is a four inter-connected loop channel type BWR with natural circulation based passive MHT cooling at all power levels. Following the recent DEC events, more emphasis is laid on very low or no radiological impact in the public domain and no requirement of evacuation in the spirit of “more good than harm”. New NPPs like NCABWR come closer to these requirements. With the mostly passive safety systems, NCABWR can avoid and mitigate DEC. The MHT system configuration for such NPPs need to be selected optimally based on the safety enhancement. In this study, optimal NCABWR MHT configuration has been further evolved based on all the Postulated Initiating Events (PIEs) including multiple failures, such as Anticipated Transients without Scram (ATWS), 200% inlet header break Loss of Coolant Accident (LOCA) without reactor trip etc. Several MHT configurations were conceptualized based on comprehensive overall design, layout requirements and reactor specific PIEs like a LOCA with a postulated break in the four partitioned ECCS header compartment. LOCA in one of the four ECCS header compartment required a mandatory interconnection among all the four MHT loops through the Common Reactor Inlet Header (CRIH) without any partition. This analysis and related issues signaled towards requirement of development of an optimal MHT configuration. The objective of the present study is to configure the MHT components and the ECCS in such an optimal formation so as to provide an enhanced performance in terms of heat removal during adverse conditions. This paper brings out clearly the important aspects needed to be considered while selecting the MHT configuration for the NPPs with enhanced safety even for ATWS-DEC scenarios, without compromising the ease of operation. The development of advanced accident tolerant fuels with improved performance, an improved/optimal MHT NPP configuration is inevitable to ensure safety.

Flux-driven integrated modelling of main ion pressure and trace tungsten transport in ASDEX Upgrade

Neoclassical and turbulent heavy impurity transport in tokamak core plasmas are determined by main ion temperature, density and toroidal rotation profiles. Thus, in order to understand and prevent experimental behaviour of W accumulation, flux-driven integrated modelling of main ion heat and particle transport over multiple confinement times is a vital prerequisite. For the first time, the quasilinear gyrokinetic code QuaLiKiz is applied for successful predictions of core kinetic profiles in an ASDEX Upgrade H-mode discharge in the turbulence dominated region within the integrated modelling suite JETTO. Neoclassical contributions are calculated by NCLASS; auxiliary heat and particle deposition profiles due to NBI and ECRH are prescribed from previous analysis with TRANSP. Turbulent and neoclassical contributions are insufficient in explaining main ion heat and particle transport inside the q = 1 surface, necessitating the prescription of further transport coefficients to mimic the impact of MHD activity on central transport. The ion to electron temperature ratio at the simulation boundary at stabilizes ion scale modes while destabilizing ETG modes when significantly exceeding unity. Careful analysis of experimental measurements using Gaussian process regression techniques is carried out to explore reasonable uncertainties. In following trace W impurity transport simulations performed with additionally NEO, neoclassical transport under consideration of poloidal asymmetries alone is found to be insufficient to establish hollow central W density profiles. Reproduction of these conditions measured experimentally is found possible only when assuming the direct impact of a saturated MHD mode on heavy impurity transport.

Performance evaluation of passive pulse generator for auto depressurization system of Advanced Heavy Water Reactor

The Advanced Heavy Water Reactor (AHWR) is a 300 MWe pressure tube type boiling light water cooled, heavy water moderated reactor. It has natural circulation based main heat transport system with integral steam drums. Recently, work has been initiated on a novel passive safety system named Passive Auto Depressurization System (PADS) for AHWR. This system is a part of shutdown cooling system and in case of occurrence of small break loss of coolant accident; PADS system mitigates chance of reactor core temperature rise. It actuates automatically at low steam drum level and rapidly depressurizes main heat transport system so that water injection from gravity driven water pool can be initiated to maintain coolant inventory for a sufficiently long time.The main component of PADS is Passive Pulse Generator (PPG), a passive device which is actuated at low steam drum level and generates high pressure signal within a short time span. The PPG generates a passive pressure pulse of around 11 bar amplitude with 8 min response time with threshold steam drum level fall of 100 mm. This pressure pulse is used for opening a passive valve which initiates decay heat removal through isolation condenser leading to depressurization of main heat transport system. An experimental facility has been designed towards the development and performance evaluation of PPG for PADS system of AHWR.This paper deals with the design of the experimental facility describing its piping layout, steam drum & heat exchanger vessel and associated instrumentation & control system. Test results along with RELAP5 analysis are discussed to validate PPG performance for PADS in AHWR.

Experimental Demonstration of Safety of AHWR during Stagnation Channel Break Condition in an Integral Test Loop

Detection of loss of coolant accident (LOCA) and generation of reactor trip signal for shutting down the reactor is very important for safety of a nuclear reactor. Large break LOCA (LBLOCA) is a typical design basis accident in all reactors and has attracted attention of the reactor designers. However, studies reveal that small break loss of coolant accident (SBLOCA) can be more severe as it is difficult to detect with conventional methods to generate reactor trip. SBLOCA in channel-type reactors is essential to consider as it may create stagnation channel conditions in the reactor coolant channel, which may lead to fuel failure, if the reactor is not tripped. Advanced heavy water reactor (AHWR) is a channel-type boiling water reactor, which may experience stagnation channel conditions in case of SBLOCA in feeder pipes. For initiating the trip signals and safe shut down of the reactor in such cases, a novel system comprising of acoustic-based sensors is incorporated in the reactor design. The system detects the peculiar sound of the steam leaked from the main heat transport system (MHTS) and generates reactor trip signal. The experimental demonstration of such new system is essential before its introduction in the reactor. The experimental demonstration of the stagnation channel break, its detection by acoustic-based sensors system, and reactor trip followed by generation of reactor trip signal was performed and presented in the paper. The experiment showed that the trip signal for AHWR can be generated within 5 s with acoustic sensor and 2 s by low flow signal and reactor trip can be ensured in 7 s following a LOCA.

Investigations of natural convection and circulation in Passive Moderator Cooling System of an advanced reactor in a scaled test facility

Passive systems are being examined for the future advanced nuclear reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator passively in case of an extended Station Black Out (SBO) condition. The heavy-water moderator gets heated due to heat transferred from the main heat transport system and the core structures; and rises upward due to buoyancy. This is cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The heat exchanger, in turn, is cooled by water from gravity driven water pool (secondary loop). The natural circulation loops provide sufficient cooling to prevent the increase of moderator temperature and pressure inside Calandria vessel beyond safe limits. In addition to this, decay heat is also dumped into GDWP tank through submerged isolation condensers.A scaled test facility was setup to simulate the thermal hydraulic characteristics of the PMCS and to analyze the performance behavior of the PMCS. Prolonged SBO transient experiment was performed in the test facility and analyzed using system code RELAP5/Mod 3.2. Besides, CFD simulations were also carried out for analyzing the temperature distribution inside calandria vessel of the test facility. Experimental results show efficient removal of the heat from the moderator and isolation condensers. The simulation results show that RELAP5 is able to predict the natural circulation phenomena very well and three dimensional temperature fields inside the vessel can be predicted by CFD. Temperature field as predicted by CFD is found to be in good agreement with the experiment.

RELAP5 simulation of CANDU Station Blackout accidents with/without water make-up to the steam generators

In the event of a complete Station Blackout (SBO) of a CANDU reactor, the current safety practice is to initiate depressurization early in the transient once a lack of power has been declared. This requires operator actions to manually open the main steam safety valves (MSSVs) on the secondary side initiating a crash-cool procedure. The depressurization of the secondary side allows make-up water to be supplied to the steam generator (SG) secondary-side thereby extending natural-circulation driven heat removal from the fuel in the primary heat transport system. Once depressurized there are several additional water sources available to replenish the secondary-side inventory, and in the event that emergency power cannot be restored emergency mitigating equipment (EME) are available to provide alternative water make-up. The objective of this paper is to examine the processes and phenomena involved during and after crash-cooling and compare these results to cases where operator actions are not credited. Simulations are performed until such time as the secondary-side inventory is stabilized from alternative water sources or until it is depleted.A detailed RELAP5 model of a 900MW CANDU plant has been created including the primary heat transport system (PHTS), the feed and bleed system, the steam generator secondary side, the moderator system, and the shield water cooling system. The 480 fuel channels were grouped into 20 channels by elevation and channel power. The models were benchmarked against the 1993 loss-of-flow event at Darlington NGS and agreed with the station data within the reported measurement uncertainty. Then the models were used to simulate the Station Blackout accidents with loss of class IV, class III, and emergency power supplies. Five different scenarios with/without crash-cool and with different water make-up options are modeled and key sensitivities determined. The results show that the depressurization of the secondary side may create a situation where continuous natural circulation breaks down and intermittent buoyancy induced flows (IBIF) takes place. The RELAP5 predicted IBIF phenomena are discussed, as well as the limitations of the current RELAP5 code. The main focus of this paper is on the early stage of the accidents, i.e. when adequate steam generator secondary side inventory exists and where damage to the main heat transport system can be precluded. The results demonstrate that EME actions to maintain SG inventory are effective and ensure fuel and fuel channel integrity.

Demonstration of adequacy of passive in-bundle ECC injection of AHWR

Emergency Core Cooling System (ECCS) of a reactor is designed for postulated Loss-of-Coolant-Accident (LOCA). Such an accident may occur due to a leakage or breakage in the main heat transport system of the reactor which is beyond the capacity of make-up systems. In Advanced Heavy Water Reactor (AHWR), in-bundle injection of emergency coolant is proposed. Unlike conventional header/plenum injection systems (as in PWRs and PHWRs) or top spray system (as in BWRs), in-bundle injection aims to deliver the emergency coolant directly to the fuel cluster in case of a LOCA. Though very promising, this concept has not been widely used in water cooled reactors. This necessitates conducting experiments for establishing the efficacy of this method of injection. The paper presents the results of such experiments on in-bundle ECC in a fuel cluster simulating the AHWR fuel cluster.

Experimental and computational simulation of thermal stratification in large pools with immersed condenser

This paper deals with experimental and numerical investigations on thermal stratification phenomenon conducted in Integral Test Loop (ITL) scaled test facility for Advanced Heavy Water Reactor (AHWR). Experiments at decay power level were designed to simulate thermal hydraulic performance of Isolation Condenser (IC) system of AHWR at prototypical pressure (70bar) and temperature (285°C). The experiments were simulated using the Best Estimate (BE) thermal hydraulic system code RELAP5. Two different pool side nodalizations were tried for simulation of IC-pool. Nodalization, where the IC-pool was simulated with one lumped volume did not simulate the observed thermal stratification behavior appropriately. However, it simulated the Main Heat Transport System (MHTS) behavior well. To simulate pool side stratification a three volume pool nodalization was adopted, which simulated the observed thermal stratification behavior in the pool well apart from simulating MHTS system behavior. Study shows a characteristic thermal stratification behavior in pool, which prevails for the initial few hours of IC functioning. The characteristic curve showing various stages of observed pool thermal stratification development and decay during experiment has been presented in the paper. An interrelation between pressure and water level into the pool has been observed in studies.

CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria. Next, a pseudo transient simulation was carried out by taking the power and mass flow rate at different time instants to the Calandria from the 1D analysis, as an input to the 3D steady state simulations. The results indicate that temperature hotspots are located at the top center of the fuel channel matrix depending on the operating regime. Furthermore, it was observed that for a period of 7 days the temperatures are well within the safe limits.

Design and development of innovative passive valves for Nuclear Power Plant applications

The recent Fukushima accident has resulted in an increased need for passive safety systems in upcoming advanced reactors. In order to enhance the global contribution and acceptability of nuclear energy, proven evidence is required to show that it is not only green but also safe, in case of extreme natural events. To achieve and establish this fact, we need to design, demonstrate and incorporate reliable ‘passive safety systems’ in our advanced reactor designs.In Nuclear Power Plants (NPPs), the use of passive safety systems such as accumulators, condensing and evaporative heat exchangers and gravity driven cooling systems provide enhanced safety and reliability. In addition, they eliminate the huge costs associated with the installation, maintenance and operation of active safety systems that require multiple pumps with independent and redundant electric power supplies. As a result, passive safety systems are preferred for numerous advanced reactor concepts. In current NPPs, passive safety systems which are not participating in day to day operation, are kept isolated, and require a signal and external energy source to open the valve. It is proposed to replace these valves by passive components and devices such as self-acting valves, rupture disks, etc.Some of these innovative passive valves, which do not require external power, have been recently designed, developed and tested at rated conditions. These valves are proposed to be used for various passive safety systems of an upcoming Nuclear Power Plant being designed by India. For example, the Hot Shutdown Passive Valves (HSPV), developed for the decay heat removal system keep the main heat transport system under hot conditions by passively sensing and controlling the system pressure. Another crucial and important valve which has been successfully developed is the Poison Injection Passive Valve (PIPV) for the Passive Poison Injection System. It not only provides higher reliability, but also ensures safe shutdown of the reactor in case of insider threats or malevolent acts in disabling active shutdown system of the reactor.Recently, an innovative valve called the Accumulator Isolation Passive Valve (AIPV) has been developed for the Emergency Core Cooling System (ECCS), which is engineered to mitigate the consequences of Loss of Coolant Accident (LOCA). During normal operation of the reactor, the pressurized accumulators (55bar) are kept isolated from the reactor core (70bar) by means of AIPVs. In case of a LOCA, these passive valves open when the main heat transport system pressure falls to a desired value. For prolonged cooling of the core, these passive valves regulate the discharge in a desired manner. These are non-standard, high pressure and high temperature valves, which are unavailable commercially and hence have to be indigenously designed and developed.This paper primarily deals with the design, development and testing of Accumulator Isolation Passive Valves (AIPV) proposed to be used in the ECCS. A 25NB size AIPV has been designed and successfully tested at Integral Test Loop (ITL) under simulated reactor conditions. It is a self-acting, ANSI 600 rating valve, which requires no external energy (i.e., neither air nor electrical power). It not only provides passive isolation but also passively controls high pressure liquid discharge through it. The design concept of the valve, functional performance, in situ valve testing methodology and the test results at simulated conditions are discussed.

Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used) is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA). In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the accident progression consequences up to a certain point. The code assesses the fuel bundle damage progression, but cannot assess further core damage for a CANDU type core, and starting from these data the amount of damaged fuel can be calculated. The radio nuclei present in the damaged fuel are supposed to be released into the main heat transport system and after that into the containment building in the worst case scenario. Assessing the radioactive nuclei maximum release is the purpose of the present paper. The radioactive nuclei release is needed for the accident management plan, limiting the environmental and population impact of the supposed accident, and furthermore for a later site remediation plan that can be put in action after the complete mitigation of the accident consequences. The maximum quantity of radio nuclei released during the accident calculated in this paper is a worst case scenario evaluation that can lead to better preparedness in an accident scenario.

Steam drum level control studies of a natural circulation multi loop reactor

Abstract The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory.

Technical feasibility study for coupling a desalination plant to an Advanced Heavy Water Reactor

Advanced Heavy Water Reactor (AHWR) is the latest Indian design for a next generation nuclear reactor. It is 300MWe, vertical, pressure tube type, boiling light water cooled natural circulation and heavy water moderated reactor. It burns thorium in its fuel core to meet the objectives of using thorium fuel cycles for commercial power generation. Desalination systems are included in the reactor circuit to produce high quality water from seawater utilizing thermal energy of AHWR to meet the reactor make-up and other process water requirements. Studies have been carried out for various options of using thermal energy of AHWR for desalination. LTE and MED-TVC desalination plants are planned to be coupled with the reactor using Main Heat Transport purification circuit waste heat and using steam from a tapping between the cross-over line of HP and LP turbine respectively. Metallic barriers, intermediate loops and pressure reversal concepts are applied to prevent radioactive ingress to product water of desalination plant. This paper describes the technical feasibility studies of coupling 250m3/day LTE and 2400m3/day MED-TVC seawater desalination plants coupled to AHWR. Coupling of desalination plants, methodologies, loss of reactor power and various safety considerations are discussed in details.

SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS

This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

Heat transport dynamics at a sandy intertidal zone

[1] Intertidal zones are spatially complex and temporally dynamic environments. Coastal groundwater discharge, including submarine groundwater discharge, may provide stabilizing conditions for intertidal zone permeable sediments. In this study, we integrated detailed time series temperature observations, porewater pressure measurements, and two-dimensional electrical resistivity tomography profiles to understand the coupled hydraulic-thermal regime of a tropical sandy intertidal zone in a fringing coral reef lagoon (Rarotonga, Cook Islands). We found three heating patterns across the 15 m study transect over tidal and diel periods: (1) a highly variable thermal regime dominated by swash infiltration and changes in saturation state in the upper foreshore with net heat import into the sediment, (2) a groundwater-supported underground stable, cool region just seaward of the intertidal slope break also importing heat into the subsurface, and (3) a zone of seawater recirculation that sustained consistently warm subsurface temperatures that exported heat across the sediment-water interface. Simple calculations suggested thermal conduction as the main heat transport mechanism for the shallow intertidal sediment, but deeper and/or multidimensional groundwater flow was required to explain temperature patterns beyond 20 cm depth. Temperature differences between the distinct hydrodynamic zones of the foreshore site resulted in significant thermal gradients that persisted beyond tidal and diel periods. The thermal buffering of intertidal zones by coastal groundwater systems, both at surface seeps and in the shallow subsurface, can be responsible for thermal refugia for some coastal organisms and hotspots for biogeochemical reactions.

Control of decay heat removal through alternate system in PFBR

In the Prototype Fast Breeder Reactor (PFBR) decay heat is removed directly from the reactor core through Safety Grade Decay Heat Removal (SGDHR) circuit. Operation Grade Decay Heat Removal (OGDHR) system offers an alternate decay heat removal path using the heat removal capacity of the main heat transport system. It is preferred path during planned shutdown and also during plant upset conditions when at least one secondary sodium system, steam water system and offsite power are available. Besides removing the decay heat, this system decreases the reactor coolant temperature at a controlled rate and it is able to maintain the plant at cold shutdown condition for fuel handling or maintenance for up to 30 days. This alternate decay heat removal system of PFBR is described in the paper along with system dynamics, analysed based on in-house developed thermal hydraulics code, sequence of the system deployment, system operation and control.