Loss Of Coolant Research Articles

Asteroseismology from Space: WIRE Monitoring of the δ Sct Star θ2 Tauri

Soon after launch in March 1999, the primary science instrument onboard the Wide-Field Infrared Explorer (WIRE) satellite failed due to loss of coolant. However, it proved possible to begin an asteroseismology program using the 52-mm aperture star camera. A few bright stars were monitored with the 512x512 SITe CCD in a bandpass approximately equivalent to V + R; further details about the orbit, the detector and the raw data reduction can be found in Buzasi et al. (2000) and Buzasi (2000). We included the binary star θ2 Tauri among the targets. It is composed of an A7IV primary and an A5V secondary (P = 140.728 d). The primary is a a δ Scuti star which has been observed several times in the last twenty years. Five terms have been determined in its light curve (Breger, 1989 and references therein); Li et al. (1997) demonstrated the amplitude variability of some of these terms by comparing different campaigns.

Safety-related plasma events for ITER-FEAT

Safety-related plasma events were investigated to assess the effects on the design of ITER-FEAT. Overpower transients were calculated to estimate the maximum possible fusion power due to loss of plasma control, and the consequent damage to in-vessel components. A passive plasma shutdown during an ex-vessel loss of coolant was investigated to estimate the maximum first wall temperature and consequent hydrogen production by the beryllium/steam reaction. Scenarios of a bounding vertical displacement event were analysed to clarify the maximum in-vessel damage due to the plasma event.

Safety and Environment Assessment of ARIES-AT

ARIES-AT is a 1000 MWe conceptual fusion power plant design with a very low projected cost of electricity. The design contains many innovative features to improve both the physics and engineering performance of the system. From the safety and environmental perspective, there is greater depth to the overall analysis than in past ARIES studies. For ARIES-AT, the overall spectrum of off-normal events to be examined has been broadened. They include conventional loss of coolant and loss of flow events, an ex-vessel loss of coolant, and in-vessel off-normal events that mobilize in-vessel inventories (e.g. tritium and tokamak dust) and bypass primary confinement such as a Loss of Vacuum and an in-vessel loss of coolant with bypass. This broader examination of accidents improves the robustness of the design from the safety perspective and gives additional confidence that the facility can meet the no-evacuation requirement. We also provide a systematic assessment of the design to address key safety functions such as confinement, decay heat removal, and chemical energy control. In the area of waste management, waste is classified by both the volume of the component and its hazard. In comparison to previous ARIES designs, the overall waste volume is less because of the compact design.

Loss of Coolant Accident and Loss of Flow Accident Analysis of the ARIES-AT Design

Loss of coolant accident (LOCA) and loss of flow accident (LOFA) analysis is performed for ARIES-AT, an advanced fusion power plant design (1000 MWe). ARIES-AT employs a high performance, high temperature blanket system. It uses the high temperature SiC/SiC for structural material and LiPb for coolant-breeder. Due to the large difference between the time scale of plasma shutdown and the coolant or power loss, it is assumed that the plasma is immediately quenched at the onset of the LOCA/LOFA and the chamber components' temperature begins to rise due to the decay heat generated. A 2-D transient finite element model is established to examine the thermal behavior of the in-vessel components to determine the maximum temperature reached, the time, and duration of the peak. The model is axisymmetric in (r-z) around the reactor axis to show the details of temperature distribution in the vertical direction. The vacuum vessel is assumed adiabatic in the inboard side and radiates to the maintenance port located on the outboard side. The maximum temperature of steel in the reactor is about (600 °C - 700°C) after about 4 days from the onset of the accident. The highest temperature in the reactor is in the divertor region and it reaches ≈1050°C after about 2-3 hours. The analysis indicates that the reactor does not need any special scheme for decay heat removal.

Hydrogen release from 800 MeV proton-irradiated tungsten rods

For the Accelerator Production of Tritium Project (APT), spallation neutrons will be moderated and then absorbed in 3He gas to produce tritium. The spallation neutrons will be generated by the interaction of high energy (~1 GeV) protons with solid tungsten rods or cylinders. A byproduct of the spallation reactions is large amounts of helium and hydrogen gas generated in the rods and other structural materials. The release kinetics of these gases during various proposed off-normal scenarios involving loss of coolant and afterheat-induced rises in temperature is of particular interest to the APT Project. In addition, however, this data is of interest for fusion reactors where tungsten used in a tokamak divertor will also be exposed to neutrons. In this case, the generated protium will be accompanied by deuterium and tritium diffusing in from the plasma-facing surface. Tungsten rods irradiated with 800 MeV protons in the Los Alamos Neutron Science Center (LANCE) to high exposures have been sectioned to produce small specimens suitable for measurement of both hydrogen and helium. Hydrogen evolution was measured by subjecting the specimens to a simulated temperature ramp from ~200 to ~1200°C, similar to that expected due to a loss of coolant and subsequent afterheat. The release measurements were conducted using mass spectrometric techniques. Four release peaks at temperatures of approximately 550, 850, 1100 and 1200°C were observed, initially suggesting a number of trapping sites with different binding energies. Subsequent analysis, however, showed that the observed peaks were artifacts of the temperature heating profile, and that the release curve could be duplicated using a single trap energy of 1.4 eV.

Safety issues associated with mobilized activation products in selected APEX designs

In the advanced power extraction (APEX) project, safety and environmental concerns considered up front, as designs evolve, so that the goal of safety and environmental attractiveness is realized. Since the neutron and surface heat loads are higher in APEX designs than those in conventional fusion designs, decay heat and activation are generally higher, presenting an increased challenge when justifying the safety case. Potential first wall materials that can function adequately under higher neutron and heat loads include materials such as tungsten and molybdenum. The activation products of both these materials are radiologically hazardous and mobilizable under accident conditions. We have examined a number of APEX concepts to determine the ability of the design to remove decay heat from the plasma-facing surface during a loss of coolant and air ingress event. In this paper, we concentrate on mobilization of first wall materials during ingress events, and provide guidance to enhance the safety characteristics of APEX designs that utilize tungsten and similar high heat load materials.

Integrated safety analysis code system (ISAS) application for accident sequence analyses

In the frame of the ITER Task ‘Reference Accident Sequences’, two accident sequences have been assessed to demonstrate the effectiveness of the use of integrated safety analysis code system (ISAS). The first one is a loss of coolant event in the divertor primary heat transfer system (DV PHTS) towards the vacuum vessel containment during normal plasma burn; the second one is a loss of coolant event in the DV PHTS towards the lower vault during baking conditions. The comparison of the results obtained in standalone and coupled modes was performed. The codes used for the analyses are ATHENA for thermal-hydraulic simulation, INTRA for containment studies and NAUA for aerosol transportation. The three codes are linked together in the ISAS chain by means of files written in the Gibiane language. The ISAS application seems to be promising in order to have a powerful tool in the fusion plant accident analysis. It is able to treat different codes, developed for different fields of applications, as a unique code simulating the complete evolution of an accident sequence. ISAS permits a detailed exchange of data between the codes, otherwise impossible in standalone mode. It reduces the errors in typing exchange data and takes into account the feedback effects during the calculation. The coupling of the codes makes possible to compensate the lack of some phenomenon models in one code if they are treated in another code of the chain (i.e. radiative heat exchange between heat structures that is missing in ATHENA and foreseen in INTRA). The results obtained in the two accident analyses demonstrate that ISAS can be very useful to represent in a more coherent way the trend of an accident respecting the correct balance of the inventories. In the paper the whole evolution of a typical fusion power plant accident is analysed starting from the break in the cooling loop system, studying the effects of the loss of coolant in the containment and simulating the leakage of the aerosols outside vacuum vessel and vaults.

Safety design for ITER and fusion power reactors

The present paper outlines key design features of relevance for the safety of international thermonuclear experimental reactor (ITER) and fusion devices of reactor scale. These features support the intrinsic safety characteristics of fusion. The most important safety function is the confinement of radioactivity. The first barrier is based on the vacuum vessel (VV) and its extensions. A pressure suppression system limits the maximum pressure of the VV in the event of loss of coolant. The second barrier is the Cryostat and its extensions. Outward leakage from the second confinement can be captured in surrounding rooms where necessary. Open penetrations of the VV will be protected by at least two separate barriers such as windows or valves. Nuclear decay heat will be removed passively by the VV cooling system. Other safety systems have been identified which play an important role for both ITER and a fusion power reactor.

Steam explosion resistance of an internal core catcher of a nuclear reactor pressure vessel

This paper deals with a feasibility design of a core catcher structure that could mitigate the consequences of the core meltdown in a commercial light water reactor. This activity has been performed in the framework of the Concerted Action IVCRS (In Vessel Core Retention Strategies), financed by the European Union in order to assess the safety of the next generation of nuclear power plants. These plants will be designed to support the consequences deriving from a core meltdown accident and from the phenomena that accompany it. Steam explosion is considered as a potential risk in the hypothesis of a severe accident occurring in a Pressurised Water Reactor nuclear power plant. The loss of coolant, which can occur in the case of a pipe break, provokes the degradation of the core geometry, its coolability and then its melting. The molten core falls down in the vessel to the lower hemispherical part rapidly transferring its energy to the water remaining in the lower plenum, which vaporises. In order to mitigate the consequences of a severe accident, a core catcher device, able to contain and to cool the molten core, has been designed. The same structure is analysed here as energy dissipators to prevent the reactor pressure vessel lower head failing in the case of steam explosion. In this paper, emphasis is placed on the structural aspects of the problem. Thermal fluid dynamics is treated macroscopically. This conservative approach allows us to overcome the actual uncertainties in the heat transfers mechanism between molten core and water. The results of a simulation, conducted with a finite element code, shows that the implementation of an internal core catcher could prevent the Reactor Pressure Vessel lower head from failing.

Load-carrying capacity and crack resistance of a cladding by the Sigma-oscillating wire technique

If cracks are postulated in the ferritic base material beneath the austenitic cladding, their initiation and propagation under hypothetical loading cases is influenced by the load carrying capacity of the cladding. The toughness of the KKS-RPV cladding was assessed by means of elastic-plastic fracture mechanics methods. Sub-sized tensile and bend specimens were fabricated by reconstitution technique from broken halves of standard ISO-V Charpy specimens, representing crack extension in radial and circumferential direction. They were tested and evaluated and further analyzed with the Gurson model. For temperatures relevant to the loss of coolant and upset conditions analyzed, a sufficient toughness of the cladding in terms of J and CTOD resistance curves could be shown.

Application of code scaling applicability and uncertainty methodology to the large break loss of coolant

In the late 1980s, after completion of an extensive research program, the United States Nuclear Regulatory Commission (USNRC) amended its regulations (10CFR50.46) to allow the use of realistic physical models to analyze the loss of coolant accident (LOCA) in a light water reactors. Prior to this time, the evaluation of this accident was subject to a prescriptive set of rules (Appendix K of the regulations) requiring conservative models and assumptions to be applied simultaneously, leading to very pessimistic estimates of the impact of this accident on the reactor core. The rule change therefore promised to provide significant benefits to owners of power reactors, allowing them to increase output. In response to the rule change, a method called Code Scaling, Applicability and Uncertainty (CSAU) was developed to apply realistic methods, while properly taking into account data uncertainty, uncertainty in physical modeling and plant variability. The method was claimed to be structured, traceable, and practical, but was met with some criticism when first demonstrated. In 1996, the USNRC approved a methodology, based on CSAU, developed by a group led by Westinghouse. The lessons learned in this application of CSAU will be summarized. Some of the issues raised concerning the validity and completeness of the CSAU methodology will also be discussed.

Investigation of oxidation resistance of carbon based first-wall liner material Aerolor AO5

Minimizing the consequences of loss of vacuum and loss of coolant into vacuum accidents with respect to fusion reactor design requires the choice of materials as oxidation resistant as possible. In continuation to the testing of carbon candidate materials under oxidizing conditions (oxygen and steam) in collaboration with NET the carbon compound Aerolor AO5 of Carbone Lorraine has been examined in detail in the test facilities INDEX2 and SPALEX at Forschungszentrum Jülich. The parameters for these oxidation experiments in the in-pore diffusion controlled regime are temperature (from 973 up to 1173 K for oxygen, from 1173 up to 1423 K for steam) and reaction gas concentration (1–20 vol.% oxygen in argon, 5–100 vol.% steam in argon). The deducted equations for the primary oxidation reactions consider the dependence of the reaction rate on temperature, partial pressure and burn-off of the material.

The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

This paper summarizes the results from the EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10 −6 yr −1, even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 −6 yr −1 and the contribution of seismic events is 1.7 10 −5 yr −1. Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability.

Conceptual Design of a 50-MW Severe-Accident–Free HTR and the Related Test Program of the HTTR

The severe-accident-free high-temperature gas-cooled reactor (SFHTR) is a prototype design for a next generation reactor. It is suitable for widespread deployment by virtue of its inherent safety features and very long refueling interval. Furthermore, its inherent safety features can be demonstrated by full-scale tests. Many of these features may be demonstrated in the High-Temperature Engineering Test Reactor (HTTR).The SFHTR is designed to have the probability of a severe accident at least two orders lower than existing systems. The fuel will not exceed its failure temperature even in the event of complete loss of coolant or complete withdrawal of two control rods. A unique configuration of burnable poisons allows a fuel cycle of 16 yr and a burnup exceeding 120 GWd/t. This feature promises very high availability and good economics.We have designed two SFHTR systems. The larger one, called the MSFHTR, has a 450- to 600-MW thermal capacity and is intended for the production of hydrogen and electricity. The smaller SFHTR (SSFHTR) is intended for remote areas, off the electrical grid, for simultaneous production of electricity and desalinated water. The SSFHTR can produce 23.5 MW(electric) plus 40 t/h of water with a net efficiency of 47%.The HTTR is capable of conducting full-scale simulation testing of key SFHTR design features in order to confirm and extend the designs and as a first step in convincing the public and the licensing authorities of the validity of demonstrable inherent safety. Design features of a 50-MW SFHTR focusing on the safety concept, safety evaluation, and core design are described. In addition, an HTTR-based test-and-development program for the SFHTR is presented.

ITER reference accidents

Abstract Reference accidents are selected to envelope possible accident sequences to confirm the adequacy of ITER's safety design. Analyses of ITER reference accidents were carried out starting from postulated initiating events (PIE) to all consequential failures including environmental releases, if any. In case of plasma control malfunctioning, up to 3.3 GW of fusion power can be reached transiently, however, no radiological consequence is expected from those transients. Assuming multiple first wall pipe failures, the pressure suppression system limits the maximum pressure within the vacuum vessel to below 0.5 MPa. Long-term decay heat removal is assured by the vacuum vessel cooling system operating in natural circulation. In case of ex-vessel coolant loss with continued plasma burn, the plasma facing components can potentially reach large temperatures. This is avoided by the fusion power shutdown system that terminates plasma burn. A small in-vessel coolant leak with failure of one vacuum vessel penetration line and several confinement options are investigated to show compliance with release limits. Failures in the magnet system are analyzed and their impact on other systems such as damage to neighboring confinement barriers are investigated by postulating air, water and helium ingress into the cryostat. Failures in the tritium plant are analyzed to demonstrate the adequacy of confinement. Radioactive releases for all reference accidents are shown to be well below conservatively defined design guidelines.

Manufacturing and Testing of Actively Cooled Test Limiters for TEXTOR Made of an LPPS Tungsten-Coated TZM Heat Sink

To investigate the erosion and redeposition phenomena of fusion-related materials under stationary conditions, actively cooled test limiters were developed for the Tokamak Experiment for Technology Oriented Research (TEXTOR). The test limiters allow experiments under stationary conditions within a plasma pulse length of 10 s. Heat loads of typically 10 MW/m2 are removed by pressurized water; the volume flow is 10 m3/h, the pressure is 15 bar, and the minimum coefficient of heat transfer is nearly 70 000 W/m2·K. The limiters were manufactured as low-pressure plasma-spraying thermally sprayed tungsten-coated heat sinks made of the molybdenum alloy TZM. The required properties of the tungsten coating were developed by the use of a statistically based optimization routine. Optimized, actively cooled limiters were successfully tested in Forschungszentrum Jülich’s Material Research Ion Beam Test Facility (MARION) with hydrogen beams. Maximum heat loads of up to ~17 MW/m2 were applied without any failure of either the heat sink or the cooling system. The steady state of the surface temperature was measured within 2 s. Analytical and numerical models describing the effects of heat load distribution and spatial temperatures were found to be in excellent agreement with numerical predictions. In an additional experiment, loss of coolant was simulated. Transition boiling was generated, and after repeated heat loads higher than 10 MW/m2, cavitational damage of the heat sink occurred. Concerning the material selection for heat sinks of hypervapotrons and other cooling systems based on enhanced boiling of the cooling liquid, this result might be of special interest.

Flooding of a large, passive, pressure-tube light water reactor

A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. Two other considerations are important. The thermal shock experienced by the calandria and pressure tubes has been evaluated and shown to be within acceptable bounds, while simultaneously cooling the graphite fuel matrix. This has been demonstrated for the bounding scenarios of sustained film boiling and rapid quenching to yield nucleate boiling on the calandria tube outer surface. Finally, although complete flooding renders the reactor deeply subcritical, various steam/water densities can be hypothesized to be present during the flooding process which could cause reactivity to increase from the initially voided calandria case. One such hypothesis which leads to the maximum possible density of the steam/water mixture in the still unflooded calandria space is entrainment from the free surface. It is shown that the steam/water mixture density yielding the maximum reactivity peak cannot be achieved by entrainment because it exceeds thermohydraulically attainable densities of steam/water by an order of magnitude.

Inherently safe pool-type reactor as a generator of low-grade heat for district heating, air conditioning and salt water desalination

The society has a heavy demand for low-grade heat to satisfy its various needs. Different factors govern the expediency of applying nuclear reactors for these purposes. The required capacity of heat sources varies in a very wide range. In a majority of cases heat sources have to be located in the immediate vicinity of the users, therefore, nuclear reactors to be used for heat generation must feature enhanced safety. Pool-type reactors can be successfully used for producing low-grade heat. Owing to their design they feature a very high safety level. The absence of positive pressure excludes the possibility of a sudden rupture of reactor tank (vessel) or a fast loss of coolant. The availability of a large amount of water in the tank ensures long-term accumulation of residual heat. The adopted integral layout of equipment, as well as natural circulation of primary coolant improve reactor reliability and safety even further. Negative temperature coefficients of reactivity provide for reactor self-protection against reactivity accidents. Pool-type reactors can be used in newly established heat supply systems and can be built in the operation systems as well, which allows to reduce fossil fuel consumption by 80–90% depending on local conditions. Pool-type reactor heat can be used for desalinating salt water and for cooling water in absorption refrigerating machines with subsequent utilization of cold water for air conditioning, cooling of special premises, and the like. Pool-type reactors can also generate electric power to their in-house needs as well as household power requirements of a neighboring town.

Fusion Safety Research and Development in JAERI

This paper describes the current status and future plans of the fusion safety research and development regarding to the developments of the dust removal system and safety analysis code and the thermofluid experiments in the Japan Atomic Energy Research Institute (JAERI) for a fusion experimental reactor. The containment of the radioactive material is the key to achieve fusion safety. In the event of accidents, the source terms need to be evaluated with sufficient accuracy. Therefore, in JAERI, the dust characterization have been investigated and the dust removal system using electric force has been developed and tested. A safety analysis code including both thermal and plasma transient analyses under the various event sequences has been developed. Moreover, the preliminary experiments of thermofluid transients in the vacuum vessel such as Ingress of Coolant Event (ICE) and Loss of Vacuum Event (LOVA) have been started and the experimental results using preliminary LOVA/ICE apparatus during 1995–1996 are summarized in this paper.

Detection and Sizing of Defects in Control Rod Drive Mechanism Penetrations Using Eddy Current and Ultrasonics

Over the last two years, concern has been generated about the capabilities of performing nondestructive evaluation (NDE) of the closure-head penetrations in nuclear-reactor pressure vessels. These penetrations are primarily for instrumentation and control rod drive mechanisms (CRDMs) and are usually thick-walled Inconel tubes, which are shrink-fttted into the steel closure head. The penetrations are then welded between the outside surface of the penetration and the inside surface of the closure head. Stress corrosion cracks initiating at the inner surface of the penetration have been reported at several plants. Through-wall cracks in the CRDM penetration or CRDM weld could lead to loss of coolant in the reactor vessel. The CRDM penetration presents a complex inspection geometry for conventional NDE techniques. A thermal sleeve, through which pass the mechanical linkages for operating the control rods, is inserted into the penetration in such a way that only a small annulus (nominally 3 mm) exists between the thermal sleeve and inside surface of the penetration. Ultrasonic (UT) and eddy current testing (ET) techniques that could be used to provide defect detection and sizing capability were investigated. NDE techniques that could be applied inside and outside the annulus were investigated, but the major goal was to provide high defect-detection sensitivity without requiring removal of the thermal sleeve. As a result of this investigation, both ET and UT techniques for the CRDM penetrations have been developed and evaluated. Long, thin probes were designed to fit into the annulus to carry both eddy current coils and irrigated ultrasonic transducers into the region of interest. The eddy current probes were used primarily to detect cracks in the penetration while the ultrasonic transducers were used to provide an estimate of the remaining wall thickness. This paper describes the ET and UT techniques, the probes developed, and the results obtained using these probes and techniques on CRDM penetration mock-ups.