Coolant Loop Research Articles

ITER divertor ex-vessel pipe break

This is a study of a divertor (DV) ex-vessel Loss Of Coolant Accident (LOCA), performed for the ITER reactor. One base case and two parameter studies are performed. The calculations are performed with the fusion MELCOR code. The initiating event, for all three analyses, is a double-ended pipe break in the DV cooling loop, which is postulated to occur in a large diameter pipe at the inlet of the main pump. The break flow is discharged into the Tokamak Cooling Water System (TCWS) vault. This is followed by plasma disruption and a break of the DV heat structure cooling pipes, discharging coolant into the Vacuum Vessel (VV). The results show, that the pressures and temperatures are kept below design limits. The results also show that activated corrosion products (ACP) and tritium are released into the TCWS vault from the coolant water, and that VV dust does not enter into the DV coolant loop, and thus does not reach the TCWS vault.

Leak Before Break application for Primary Coolant Loop and Surge Line of VVER-1000/320 Plant: TACIS Project R2.09/96

Activities were carried out within the frame of TACIS-96 Program—Project R2.09/96 named “LBB Applicability Review and Basic Implementation Engineering for Primary Coolant Loop & Surge Line of VVER-1000/320 Nuclear Power Plant”. Main objective of the Project was to perform a Leak Before Break assessment for the Main Coolant Pipes and Pressurizer Surge Line of the Reactor Coolant System of a selected “reference” Unit of a VVER-1000, type V-320 Nuclear Power Plant, which was assumed the Balakovo Unit 2. This has included the following: • definition of a “reference” methodology for primary piping systems with cladding; • verification of the Leak Before Break concept applicability to primary piping, in accordance with the “reference” methodology defined; • preparation of necessary materials data and validation files; • evaluation of effectiveness and reliability of the Leak Detection systems and the In-service Inspection Programs (methodologies and qualification); • preparation of a recommendations list to improve the existing information and diagnostic systems; • transfer of know–how from Consortium to Sub-Contractor, integrating the feedback of Western experience. One of the main highlights of the Project was the definition and execution of a comprehensive Material Testing Program in order to establish the actual tensile and fracture properties of the primary piping materials: base materials, weld metals, and “heat affected zones” were tested. Compact tests and standard specimens for J– R curves determination were used; special attention was paid to the dissimilar typical joints of the Surge Line pipes, being made by carbon steel base metal with stainless steel weld. Fracture properties as J– R curves were determined both at room and operating temperatures. The “reference” methodology defined has taken into account both Western and Russian Leak Before Break approaches. The analysis has shown that criteria for Leak Before Break concept applicability to the primary piping of Balakovo VVER-1000/320 Unit 2, presented in the reference methods developed in the frame of this Project, are met.

Magnetic Properties at Footpoints of Hot and Cool Loops

Observations of the solar corona with Yohkoh, the Solar and Heliospheric Observatory, and the Transition Region and Coronal Explorer (TRACE) have revealed that individual coronal loops of active regions have their own temperatures from 1 to 5 MK. The hot (2-5 MK) Soft X-Ray Telescope (SXT) loops appear to require more heating energy than the cool (1-2 MK) EUV loops. We investigate the photospheric magnetic signature for the hot and cool loops with the Advanced Stokes Polarimeter. In contrast to the cool loops, the hot loops observed with the SXT are usually diffuse, resulting in ambiguous identification of their footpoint locations. We use TRACE moss structure, which we confirm is low-lying EUV emission at the footpoints of the hot loops. Footpoints of both loops have magnetic fields whose strength is 1.2-1.3 kG, and the orientation is almost vertical to the surface. A significant difference is discovered in the magnetic filling factor, which is defined by the fraction of a pixel filled with a magnetized atmosphere. The footpoints of the hot loops have a lower filling factor than the footpoints of the cool loops. We suggest that braiding of coronal magnetic fields is more efficient at the footpoints of the hot loops than at the footpoints of the cool loops as a result of the combination of the lower filling factor and higher horizontal velocity.

Analysis of an energy recovery system for reformate-based PEM fuel cells involving a binary two-phase mixture

A comprehensive analysis on a novel energy recovery system for reformate-based proton exchange membrane (PEM) fuel cell systems is presented. The energy recovery system includes a throttling valve, a heat exchanger, a compressor, and is coupled with a coolant loop for the fuel cell stack. The feed stock of the fuel reformer, which is primarily a mixture of water and fuel, is vaporized in the heat exchanger and is then compressed to a sufficiently high pressure before it is ducted into the fuel reformer. The analysis includes the throttling of two-phase fuel/water mixture and vaporization in the heat exchanger to obtain the temperature and pressure of the mixture at the inlet of the compressor. The results indicate that the power plant efficiency with the energy recovery system can be increased by more than 20% compared to that of a fuel cell power plant without the energy recovery system. Additionally, more than 25% of the waste heat generated by the fuel cell stack can be removed due to the energy recovery system, and the fuel burned for the fuel reforming purpose is reduced by more than 70%.

Boric acid recovery using polymer filtration: Studies with alkyl monool, diol, and triol containing polyethylenimines

Three water-soluble polymers containing linear alkyl monool, 1,2-diol, and 1,2,3-triol groups, mostly on the primary amines of polyethylenimine, were synthesized, characterized, and tested for their ability to recover boric acid. The boron-binding capacities of these polymers and the backbone polyethylenimine were determined by titration, ultrafiltration, and inductively coupled plasma/atomic emission spectroscopy analysis. At low boron concentrations, the 1,2,3-triol polymer performed better than the 1,2-diol, whereas at high boron concentrations, the 1,2-diol outperformed the 1,2,3-triol. 11B-NMR spectroscopy and retention studies with various salt concentrations indicated that boron interacted with these two polymers by means of ion pairing with the protonated amines and by borate ester formation. For the monool and the polyethylenimine backbone, the mechanism for boron binding was ion pairing only. These polymers are under consideration for the selective recovery and recycling of enriched boric acid used in the primary coolant loop of pressurized water nuclear reactors. © 2005 Wiley Periodicals, Inc. J Appl Polym Sci 97: 1590–1604, 2005

Analysis of the Polonium Hazard in Nuclear Power Systems with Lead–Bismuth Coolant

Polonium presents no danger under normal operating conditions when the coolant loop is sealed. In our country, we have almost 40 years of experience in working with radioactive lead–bismuth coolant in the reactors used in nuclear-powered submarines and ground-based testing prototypes. This experience has made it possible to study how the polonium radiation conditions inside working enclosures are produced and to develop effective safety measures for different operating regimes, including irregular situations and accidents. Data are presented on the polonium radiation conditions in certain typical operating and accident situations in a nuclear submarine: spillage of a substantial amount of coolant directly into the reactor bay, the consequences of an interloop leak in a steam generator, and refueling with fresh nuclear fuel. Regular medical-biological examinations of the workers participating in the operation of the system, repair work, and liquidation of the consequences of accidents did not show any cases of irradiation by polonium above the established health norms. Inferences are drawn from studies performed in this and other countries of polonium emission from lead-based polonium-containing media when such media are heated in vacuum and in an inert-gas atmosphere in the temperature range 300–800°C.

Soft X-ray analysis of a loop flare on the Sun

We present the results of an analysis of soft X-ray images for a solar flare which occurred on 1992 July 11. This flare, as seen in Yohkoh Soft X-ray Telescope (SXT) images was of comparatively simple geometry, consisting of two bright footpoints early in the flare with a bright loop seen later in the flare. We examine how closely this flare compares with the supposed paradigm of a confined simple-loop flare. Closer examination of the SXT images reveals that the flare structure consisted of at least two adjacent loops, one much fainter than the other. We examine the brighter of the two soft X-ray loops. The SXT images reveal an apparent slow, northward motion of this loop (roughly transverse to its major axis). Examination of derived emission measure and temperature images also indicate an apparent northward motion. In addition, we find an increase in the cross-sectional width at the top of the loop with time. Emission measure maps derived from the SXT images also indicates an apparent broadening of the loop-top region. We infer that the apparent northward motion and the apparent broadening of the soft X-ray emission can be explained in a reconnection scenario where successive magnetic field structures do not lie in a plane but are tilted to the south of the line of sight but with successively brightening loops oriented at less tilted angles. Halpha images for this flare reveal an evolution from a few brilliant points to a short two- ribbon-like appearance. Comparison of the SXT images with the Halpha images shows that the Halpha patches are aligned with the footpoints of the soft X-ray loops, suggesting the presence of a small arcade structure. There is no clear evidence for an eruptive signature in our observations nor in reports from other observations. The lack of an eruptive signature could suggest that the flare may have been a confined simple-loop flare, but this is not compelling due to a gap in the coronal observations prior to and early in the event. Analysis of our observations indicate that the flare exhibited characteristics suggesting that it may be better understood as a mini-arcade flare. These results casts doubt on the validity of the supposed paradigm of a confined simple-loop flare, at least for this flare. They indicate that even an apparently simple-loop flare may be considered to be a variety of arcade flare. We also find an effect which, to our knowledge, has not been reported before: the hot flaring regions later become cooler than the surrounding quiescent corona. That is, the flare loops do not evolve into bright active region loops, but into cooler loops. This may indicate an increase in the efficiency of the cooling mechanism or a transformed equilibrium state within the flaring loops.

Thermal-mixing analyses for safety injection at partial loop stagnation of a nuclear power plant

When a cold HPSI (High Pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (Pressurized Thermal Shock) evaluation be performed. Also, previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.

Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold emergency core cooling (ECC) water entering the RPV through the ECC injection into the cold leg. The experimental results show an incomplete mixing with typical concentration and temperature distributions at the core inlet, which strongly depend on the boundary conditions. Computational fluid dynamics calculations were found to be in good agreement with the experiments.

Spatial and Temporal Properties of Hot and Cool Coronal Loops

A suite of images from the XUV Doppler Telescope (XDT), the Yohkoh Soft X-ray Telescope (SXT), and the Extreme-Ultraviolet Imaging Telescope (EIT) on the Solar and Heliospheric Observatory (SOHO) allow us to see the whole (T > 1 MK) temperature evolution of coronal loops. The detailed morphological comparison of an active region shows that hot loops seen in SXT (T > 3 MK) and cool loops seen in the the EIT 195 A band (T ~ 1.5 MK) are located in almost alternating manner. The anticoincidence of the hot and the cool loops is conserved for a duration much longer than the estimated cooling timescale. However, both hot and cool loops have counterparts in the intermediate-temperature images. The cross-correlation coefficients are higher for neighboring temperature pairs and lower for pairs with larger temperature differences. These results suggest that loops are not isothermal but rather have a differential emission measure distribution of modest but finite width that peaks at different temperatures for different loops.

Reactor Chamber and Balance-of-Plant Characteristics for a Fast-Ignition Heavy-Ion Fusion Power Plant

The concept of a fast-ignition heavy-ion fusion (FIHIF) power plant involves a cylindrical target and superhigh energy ion beams. The driver produces one plus/minus charge state multimass platinum ions with energy of 100 GeV. The driver efficiency and the target gain are taken as 0.25 and 100, respectively. The preliminary data on the energy fluxes delivered to the reactor chamber wall by the 500-MJ fusion yield are presented. The reactor chamber designed has two sections. In the first section, the microexplosions occur, and in the second section of bigger volume the expansion and condensation of vapors take place. The response of the blanket and the thin liquid film at the first-wall surface is evaluated. Lithium-lead eutectic is taken as a coolant. The evaporated mass and the condensation time are estimated, taking into account major thermophysical effects. The estimated neutron spectrum from the FIHIF target gives an average neutron energy of 11.9 MeV. The mechanical stresses in the construction material due to neutron energy release are evaluated. The outlet coolant chamber temperature is taken as 550°C. The heat conversion system consisting of three coolant loops provides a net efficiency of the FIHIF power plant of 0.37.

A self-consistent treatment of radiation in coronal loop modelling

We perform a hydrodynamic simulation of a cooling coronal loop and calculate the time-dependent ion populations of the most abundant elements of the solar atmosphere at each time-step. We couple the time-dependent ion balance equations to the hydrodynamic equations in order to treat the energy loss through radiation in a self-consistent way by allowing for the emission from a potentially nonequilibrium ion population. We present results for the response to the changing conditions in the loop of the population of C VII ions and find significant deviations from equilibrium in the coronal and footpoint regions of the loop. The former is due to the tenuous nature of the coronal plasma causing recombinations to be rare and the latter is due to the strong downflows that develop as the loop cools, which carry persistent C VII ions into the lower regions of the loop. We also present a comparison between total plasma emissivity curves calculated during this simulation and an almost identical simulation that assumed an equilibrium ion population for the calculation of the radiation term. As a result of the nonequilibrium ion populations we find significant differences between the emissivity curves of each simulation and the loop cooling times. We suggest that a consideration of nonequilibrium ionisation and radiation might help to (a) explain the thermal broadening observed in some emission lines during explosive events, and (b) reconcile differences between theory and observations relating to the longevity of some loops observed in the TRACE filters.

Hydrogen removal from LWR containments by catalytic-coated thermal insulation elements (THINCAT)

In the THINCAT project, an alternative concept for hydrogen mitigation in a light water reactor (LWR) containment is being developed. Based on catalytic coated thermal insulation elements of the main coolant loop components, it could be considered either as an alternative to backfitting passive autocatalytic recombiner devices, or as a reinforcement of their preventive effect. The present paper summarises the results achieved at about project mid-term. Potential advantages of catalytic thermal insulation studied in the project are: • reduced risk of unintended ignition, • no work space obstruction in the containment, • no need for seismic qualification of additional equipment, • improved start-up behaviour of recombination reaction. Efforts to develop a suitable catalytic layer resulted in the identification of a coating procedure that ensures high chemical reactivity and mechanical stability. Test samples for use in forthcoming experiments with this coating were produced. Models to predict the catalytic rates were developed, validated and applied in a safety analysis study. Results show that an overall hydrogen concentration reduction can be achieved which is comparable to the reduction obtained using conventional recombiners. Existing experimental information supports the argument of a reduced ignition risk.

Reactor noise measurements in the safety and regulating systems of CANDU stations

Reactor noise measurements of safety and regulating system intrumentation are performed in the CANDU nuclear power stations of Ontario Power Generation (OPG) and Bruce Power. Station signals included in the noise measurements are in-core flux detectors (ICFD), ion chambers (I/C), flow transmitters, pressure transmitters, and resistance temperature detectors (RTD). Their frequency dependent noise signatures are regularly measured during steady-state operation, and are used for parameter estimation and anomaly detection.The specific applications include the following areas:⊎ Flux noise measurements to detect and characterize (a) anomalies of in-core flux detectors, ion chambers and their electronics, (b) mechanical vibration of fuel channels and in-core detector tubes induced by coolant/moderator flow.⊎ Pressure and flow noise measurements to estimate the in-situ response times of flow/pressure transmitters and their sensing lines installed in the reactor's coolant loops.⊎ Temperature noise measurements to estimate the in-situ response times of thermal-well or strap-on type RTDs installed in the reactor's coolant and moderator loops.

Temperature and Density Structure of Hot and Cool Loops Derived from the Analysis ofTRACEData

We study the transversal structure (in particular the filamentation) and the longitudinal plasma stratifica- tion in two sets of solar coronal loops observed with TRACE in the 171 and 195 Apassbands. The density stratification and the thermal structuring of the plasma along the fibrils that form the loops are derived using two techniques simultaneously: (1) a filter ratio diagnostic (195/171 A ˚ ) and (2) modeling intensity profiles along the fibrils with hydrostatic models; in both techniques we remove the background flux. We find consistent values of temperature and density with both methods in the coronal structures selected. We find evidence of a very warm fibril (T � 5 � 10 6 K) and of rather cold ones (T � 2 � 10 5 K). The hot fibril appears to have a thermal structure with a maximum at the apex compatible with temperature and density stratification, dictated by energy balance, typical of a nonisothermal hydrostatic loop. The cold fibrils appear to be isothermal and are probably in dynamical conditions. Subject headings: Sun: corona — Sun: UV radiation On-line material: color figures

Performance evaluation of an energy recovery system for fuel reforming of PEM fuel cell power plants

This paper describes an energy recovery system that recovers waste thermal energy from a fuel cell stack and uses it for fuel reforming purposes. The energy recovery system includes a throttling valve, a heat exchanger, and a compressor, and is coupled with a coolant loop of the fuel cell stack. The feed stock of a fuel reformer, which is primarily a mixture of water and fuel, is vaporized in the heat exchanger and is compressed to a sufficiently high pressure before it is ducted into the fuel reformer. The performance of a fuel cell power plant equipped with the energy recovery system is evaluated. The results indicate that the power plant efficiency can be increased by more than 40% compared to that of a fuel cell power plant without the energy recovery system. Additionally, up to 90% of the waste heat generated in the fuel cell stack is recovered. As a result, the required heat dissipation capacity of the radiator that is used for cooling the fuel cell stack can be drastically reduced.

Consideration for the Safety of a Fusion Power Plant.

Based on the fundamental approach for safety of ITER, a possible extension of it to assure the safety of fusion power plant was considered. Although the entire an alysis and licensing preparation are specific for ITER, its methodology which takes full advantage of the inherent features of fusion is expected to be applied to the fundamental logic of fusion power plants. Both energy and radioactive source terms that could be potential hazards are typically operative for a number of days rather than for a year, as in the case of fission. The major differences from the test reactor ITER were identified as the power blanket, coolant loop, and a generator train that will hold high temperature and a considerable amount of tritium. It is anticipated that the tritium inventory and most of the tritium plant would essentially be same as those for ITER, tritium recovery and removal from the blanket loop will dominate the fusion power plant tritium systems. Such a tritium system will actively remove tritium at a daily throughput of the order of plant inventory. This feature suggests that no dedicated off-normal systems are needed to assure the safety of the fusion plant in terms of environmental tritium release.

Structure mechanics and thermal-hydraulic behaviour of a PWR coolant loop with a postulated crack like leakage

Thermal hydraulic calculations, using ATHLET, have been used to evaluate pressures and temperatures in primary and secondary circuits, following a postulated leak in the surge line. These were, then, used as input for the structural mechanics calculations with ADINA. The main results of the analyses may be summarised as follows; global deformations are significantly reduced by the drop in pressure and decrease in temperature (e.g. the vertical displacement of the surge line in the region of the crack is reduced by 50%); the leakage area at the end of the transient is about 30% lower compared with the value at the beginning; the leak rate is slightly increased at the end of the transient in comparison with the initial rate; the consideration of the crack surface pressure is important, as it leads to a significant increase in crack load and leakage area by about 50 and 35%, respectively.

Classification trees in APL

This paper considers the problem of classification tree based data analysis. Among the topics discussed in the paper are: growing a classification tree using CART style exhaustive search for splits, selecting the right size for the tree using minimal cost-complexity cross-validation pruning, and examples of the application of classification trees.The algorithms are implemented in Dyalog APL.Application example is based on data from vibration monitoring equipment installed on a Nuclear Power Plant in Novovoronehz, Russia and includes classification of vibration spectra of steam generators or coolant pumps and classification of vibration spectra of steam generators of different coolant loops.

Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR’s surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip.