Tribological performance of MoS2 coatings in liquid helium and at high loads

The main part of a narrow support element (NSE) of the W7-X superconducting coil system is an aluminium bronze pad, PVD coated on its spherical surface with MoS₂, which slides against the flat surface of the stainless steel coil housing, coated with MoS₂ spray. The operational requirements of the NSEs are: vacuum of p < 10⁻⁶ mbar, temperature T  4 K, maximum load P 1500 kN, typical displacement ≤5 mm, smooth sliding and no stick-slip events. The paper describes test results obtained with a downscaled NSE at T = 4.2 and 77 K. During the test the NSEs were submerged in liquid helium and nitrogen, respectively. Whereas the LN₂ test ran smoothly for up to 15,000 cycles, the test in LHe showed stick-slip from the very first cycle. The stick-slip disappeared after 50 cycles. Post mortem analysis of the tested parts revealed that in case of LHe the sprayed MoS₂ film was removed during the first 30–100 cycles by blistering and flaking. The reason for the loss of adhesion at LHe temperature is not known, several possible causes are under discussion. Further experiments under vacuum and at T 4 K are being prepared which are expected to help in clarifying the issue.     

Experimental study on natural circulation and its stability in a heavy liquid metal loop

Motivated by the increasing interest in heavy liquid metal (HLM) cooled fast reactors and accelerator driven system (ADS), the TALL test facility was designed and constructed at KTH to investigate the thermal-hydraulic characteristics of HLM. In this paper, the HLM natural circulation characteristics in a HLM loop were investigated with experiments in the TALL test facility. The study includes measurements on (1) start-up of natural circulation from different initial conditions; (2) stability of natural circulation; (3) effects of influencing parameters and (4) capability of natural circulation. The experimental data are compared to predictions with a relevant code (RELAP5). Significant natural convection flow was observed in the experiments. It was found that the natural circulation was easily established and stabilized. It took only a few minutes to have a stable natural circulation prevailing from cold conditions. The natural circulation flowrate depends on the loop resistance, and the temperature difference between the hot leg and the cold leg, as determined by the power level and the heat sink capacity. The experiments show that the maximum flowrate for the natural circulation is 0.5 kg/s (corresponding to 0.5 m/s in the heat exchanger), resulting in heat removal of 15 kW from the core tank, which is comparable to the capacity of 100 W/cm of the electric heater elements. The preliminary analysis performed with the RELAP5 code is in reasonable agreement with the experimental data.     

Effects of working fluid and shaft rotation speed on the performance of HTR plants and the size of CBC turbo-machine

Investigated are the effects of the molecular weight of the working fluid, reactor exit temperature, and shaft rotation speed on the size and number of stages of the turbo-machine as well as the performance of high temperature reactor (HTR) plants with actively cooled reactor pressure vessel and direct or indirect Closed Brayton Cycles (CBCs). The present analyses for working fluids of helium (4 g/mol) and the 15 g/mol He–Xe and He–N₂ binary mixtures are performed for a reactor thermal power of 600 MW, shaft rotation speed of 3000–9000 rpm, and reactor exit temperature from 973 K to 1223 K. For the plants with indirect CBCs, the analyses assume a temperature pinch of 50 K in the IHX. Results show that the CBC compression ratio is relatively low (2.6 for He and He–Xe and 3.2 for He–N₂), increases very little with increasing the reactor exit temperature and near the maximum thermal efficiency of the plant. For the plants with a direct helium CBC, the thermal efficiency increases from 42% to 51% as the reactor exit temperature increases from 973 K to 1223 K, respectively, versus 37% to 47% for the plants with indirect He-CBC. The HTR plants with indirect He–Xe and He–N₂ CBCs and operating at a turbine inlet temperature of 1123 K have slightly higher thermal efficiencies (45.9% and 45.8%) than the He plant with indirect CBC (45.6%), generating 1.6 MWe more electrical power. The molecular weight of the working fluid has a very small effect on the plant thermal efficiency, but significantly reduces the size and number of stages of the CBC turbo-machine. Increasing the shaft rotation speed also decreases the size and number of stages of the CBC turbo-machine.     

Noble gas binary mixtures for gas-cooled reactor power plants

This paper examines the effects of using noble gases and binary mixtures as reactor coolants and direct closed Brayton cycle (CBC) working fluids on the performance of terrestrial nuclear power plants and the size of the turbo-machines. While pure helium has the best transport properties and lowest pumping power requirement of all noble gases and binary mixtures, its low molecular weight increases the number of stages of the turbo-machines. The heat transfer coefficient for a He–Xe binary mixture having a molecular weight of 15 g/mole is 7% higher than that of helium, and the number of stages in the turbo-machines is 24–30% of those for He working fluid. However, for the same piping and heat exchange components design, the loop pressure losses with He–Xe are 3 times those with He. Consequently, for the same reactor exit temperature and pressure losses in piping and heat exchange components, the higher pressure losses in the nuclear reactor decrease the net peak efficiency of the plant with He–Xe working fluid (15 g/mole) by a little more than 2% points, at higher cycle compression ratio than with He working fluid.     

Activity level of gross α and gross β in airborne aerosol samples around the Qinshan NPP

The monitoring results of gross α and gross β activity from 2001 to 2005 for environmental airborne aerosol samples around the Qinshan NPP base are presented in this paper. A total of 170 aerosol samples were collected from monitoring sites of Caichenmen village, Qinlian village, Xiajiawan village and Yangliucun village around the Qinshan NPP base. The measured specific activity of gross α and gross β are in the range of 0.02 - 0.38 mBq/m^3 and 0.10 - 1.81 mBq/m^3, respectively, with an average of 0.11 mBq/m^3 and 0.45mBq/m^3, respectively. They are lower than the average of 0.15 mBq/m^3 and 0.52mBq/m^3, of reference site at Hangzhou City. It is indicated that the specific activity of gross α and gross β for environmental aerosol samples around the Qinshan NPP base had not been increased in normal operating conditions of the NPP.     

Increased fuel burn up in a CANDU thorium reactor using weapon grade plutonium

Weapon grade plutonium is used as a booster fissile fuel material in the form of mixed ThO₂/PuO₂ fuel in a Canada Deuterium Uranium (CANDU) fuel bundle in order to assure the initial criticality at startup.Two different fuel compositions have been used: (1) 97% thoria (ThO₂) + 3%PuO₂ and (2) 92% ThO₂ + 5% UO₂ + 3% PuO₂. The latter is used to denaturize the new ²³³U fuel with ²³⁸U. The temporal variation of the criticality k_∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 20 years. The criticality starts by k_∞ = 1.48 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout. The criticality becomes quasi constant after the second year and remains above k_∞ > 1.06 for 20 years. After the second year, the CANDU reactor begins to operate practically as a thorium burner.Very high burn up could be achieved with the same fuel material (up to 500,000 MW·D/T), provided that the fuel rod claddings would be replaced periodically (after every 50,000 or 100,000 MW·D/T). The reactor criticality will be sufficient until a great fraction of the thorium fuel is burnt up. This would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.     

Neutronic performance of HYLIFE-II fusion reactor using various thorium molten salts

HYLIFE-II is one of the major inertial fusion energy reactor design concepts in which a thick molten salt layer (Flibe = Li₂BeF₄) is injected between the reaction chamber walls and the explosions. Molten salt coolant eliminates the frequent replacement of solid first wall structure during reactors lifetime by decreasing intense neutron flux. This study presents the neutronic analysis of HYLIFE-II fusion reactor using various liquid wall coolants, namely, 75% LiF–25% ThF₄, 75% LiF–24% ThF₄–1% ²³³UF₄ or 75% LiF–23% ThF₄–2% ²³³UF₄. Neutron transport calculations for the evaluation of neutron spectra were conducted with the help of Scale 4.3 by solving the Boltzmann transport equation in S₈–P₃ approximation. The effects of flowing liquid wall thickness and type of coolant on the neutronic performance of the reactor were investigated. Furthermore, radiation damage calculations at the first wall structure with respect to type and thickness of liquid wall were carried out. Numerical results showed that using the flowing liquid wall containing the molten salt, 75% LiF–23% ThF₄–2% UF₄ with a thickness of 70 cm maintained tritium self-sufficiency of the (DT) fusion driver and extended the first wall lifetime to the reactors lifetime (30 full power years). In addition significant amount of high quality fissile fuel was bred through (n, γ) reaction of ²³²Th. Moreover, energy multiplication factor (M) was increased to 12 by high rate fission reactions of ²³³U occurring in the flowing wall. On the other hand, it was concluded that using the other two coolants, 75% LiF–25% ThF₄ or 75% LiF–24% ThF₄–1% ²³³UF₄, as liquid wall did not satisfy the radiation damage and the tritium sufficiency criteria together at any thickness, so that these two coolants were not suitable to improve neutronic performance of HYLIFE-II reactor.     

Stopping power of polymeric foils for swift heavy ions

The stopping power of polypropylene PP(C₃H₆) and Polyethylene naphthalate PEN (C₇H₅O₂) polymeric foils has been measured, using transmission technique, for Si, Cl and Ti ions covering the energy range 1.0–4.5 MeV/u. These measured stopping power values have been compared with the corresponding values generated from the widely used semi-empirical formulations and standard data tables. The applicability of these formulations and data tables, in the light of the experimental values, is highlighted.     

Experimental and numerical investigation of a full-scale helium-cooled divertor finger mock-up

A modular helium-cooled divertor design based on the multi-jet impingement concept (HEMJ) that is capable of accommodating a surface heat flux of at least 10 MW/m² has been developed at the Forschungszentrum Karlsruhe. Experimental investigations with a full-scale mock-up designed and built at the Georgia Institute of Technology, Atlanta were carried out in the helium loop HEBLO at Karlsruhe. Tests were run at heat loads of up to 2 MW/m² and flow conditions of 38 °C and 8 MPa so that the Reynolds number matches that for the actual divertor operating conditions (21,600). Comparison between the experimental data and results of simulations performed using the computational fluid dynamics code ANSYS CFX showed good agreement for the cooled surface temperature distribution, while the pressure loss was underestimated by about 20% by the code.     

AgInCd control rod failure in the QUENCH-13 bundle test

The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO₂ pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI–Switzerland and AEKI–Hungary.Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g of H₂ during the pre-reflood phases). Posttest examinations of bundle structures revealed the presence of only little relocated AgInCd melt in the form of rivulets, mainly in the coolant channels surrounding the control rod simulator and at axial elevations between the third (0.55 m) and first spacer grids (−0.1 m).Results of QUENCH-13 on the onset of absorber rod failure are in agreement with CORA results of nine experiments each containing one or more AgInCd/stainless steel/Zircaloy-4 control rod assemblies. Bundle degradation triggered by early melt formation was, however, more pronounced in the CORA experiments with maximum bundle temperatures of 2300 K (compared to 1800 K in QUENCH-13). Consequently, QUENCH-13 allowed studying the initiation of absorber rod failure by eutectic reactions of SS-Zr, and later on of AgInCd-Zr, as well as the redistribution of the absorber material within the test bundle. Furthermore, input data for modeling of aerosol release during severe accidents are considered as benefits of the experiment.     

The MEGAPIE-TEST project: Supporting research and lessons learned in first-of-a-kind spallation target technology

The megawatt pilot experiment (MEGAPIE) has been launched by six European institutions (PSI, FZK, CEA, SCK-CEN, ENEA and CNRS), JAEA (Japan), DOE (US) and KAERI (Korea) with the aim to carry out an experiment, in the SINQ target location at PSI (Switzerland), to demonstrate the safe operation of a liquid metal (lead–bismuth eutectic, LBE) spallation target hit by a 1 MW proton beam. The European Commission has joined the MEGAPIE project through the 5-year (2001–2006) project named MEGAPIE-TEST. This project has been formally concluded with an International Workshop, where the results and the lessons learned during the project have been summarised. This work presents a review of the outcome of that Workshop.     

Development and steady state level experimental validation of TASS/SMR core heat transfer model for the integral reactor SMART

An advanced integral pressurized water reactor (PWR) of a small size (330 MWt) is being developed by the Korea Atomic Energy Research Institute (KAERI). The purposes of the reactor are a sea water desalination and an electricity generation. To enhance its safety, many advanced design concepts are introduced such as a passive residual removal system and a low power density core. For the safety validation of the designed reactor, a system analysis code named TASS/SMR, was developed. TASS/SMR code uses a one dimensional node/path modeling for the thermal hydraulic calculation and point kinetics for the core power calculation. The code also has specific models for the developed integral reactor, such as a helical tube heat transfer model and a passive residual heat transfer model. One of the important models for the safety or performance calculation is the core heat transfer model. The core heat transfer model of TASS/SMR was developed to meet the requirements of the 10 CFR 50 appendix K EM model as well as the realistic models. The developed model was validated with experimental data. The results show that the model predicts the heat transfer phenomena in the reactor core with a reasonable conservatism.     

Very high swelling and embrittlement observed in a Fe–18Cr–10Ni–Ti hexagonal fuel wrapper irradiated in the BOR-60 fast reactor

The highest void swelling level ever observed in an operating fast reactor component has been found after irradiation in BOR-60 with swelling in Kh18H10T (Fe–18Cr–10Ni–Ti) austenitic steel exceeding 50%. At such high swelling levels the steel has reached a terminal swelling rate of 1%/dpa after a transient that depends on both dpa rate and irradiation temperature. The transient duration at the higher irradiation temperatures is as small as 10–13 dpa depending on which face was examined. When irradiated in a fast reactor such as BOR-60 with a rather low inlet temperature, most of the swelling occurs above the core center-plane and produces a highly asymmetric swelling loop when plotted vs. dpa. Voids initially harden the alloy but as the swelling level becomes significant the elastic moduli of the alloy decreases strongly with swelling, leading to the consequence that the steel actually softens with increasing swelling. This softening occurs even as the elongation decreases as a result of void linkage during deformation. Finally, the elongation decreases to zero with further increases of swelling. This very brittle failure is known to arise from segregation of nickel to void surfaces which induces a martensitic instability leading to a zero tearing modulus and zero deformation.     

High-energy γ-irradiation effect on physical ageing in Ge–Se glasses

Effect of Co⁶⁰ γ-irradiation on physical ageing in binary Ge_xSe_100–x glasses (5  x  27) is studied using conventional differential scanning calorimetry method. It is shown, that high-energy irradiation leads to additional increase in the glass transition temperature and endothermic peak area near the glass transition region over the one induced by isochronal storage of these glasses at normal conditions. This γ-induced physical ageing is shown to be well-pronounced in Se-rich glasses (x < 20), while only negligible changes are recorded for glasses of 20  x  27 compositions. The effect under consideration is supposed to be associated with γ-activated structural relaxation of the glass network towards thermodynamic equilibrium of supercooled liquid.     

Effects of interfacial shear condition and trailing-corner radius on the wake vortex of a bubble

Comparative effects between the interfacial shear condition and the trailing-corner radius () on the wake vortex of a bubble are studied. In the investigation, the standard k– model is employed, and the two types of bubble: solid and gaseous, have different interfacial boundary condition. Namely, for solid bubbles the no-slip condition is imposed, resulting in a non-zero interfacial shear condition, while for gaseous bubbles the free-slip condition is imposed, yielding a zero interfacial shear condition. The flow condition is set for a slug flow with the bubble drifting at a terminal velocity corresponding to the Reynolds number of 35,000. The results show that, the flow can be roughly divided into two flow regimes: the small- and large- regimes. In the small- regime, the trailing-corner radius plays a dominant role and the difference in the interfacial shear condition has little effects on the wake vortex, causing the wake vortices of the two bubble types to be similar in shape, size, and circulation. In contrast, in the large- regime, the interfacial shear condition can manifest and affect flow separation and the wake vortex, causing significant differences between the wake vortices from the two bubble types. Namely, as is increased towards the large- regime, the wake vortex of the solid bubble changes relatively little while that of the gaseous bubble significantly decreases in size. At small- the circulations around the wake vortex of both types of bubble are almost identical initially. However, as is increased towards the large- regime, the circulation of the gaseous bubble decreases with increasing at a more pronounced rate than that of the solid bubble. These results show that it is the absence of interfacial shear in the large- regime that causes the wake vortex to be more sensitive to the trailing-corner radius.     

Experimental investigation on similarity between velocity and density profiles in density-stratified countercurrent flow in reactor horizontal leg

A density-stratified countercurrent flow was investigated to obtain data necessary to develop a physical model on a thermally stratified flow in a horizontal leg of a pressurized water reactor (PWR). The experiments were conducted at atmospheric pressure and temperature using fresh water and NaCl solution with a non-dimensional density ratio of up to 1.2. The emphasis was placed on measurements of velocity and concentration profiles near the interface between the two fluid layers. Measured mean velocity and concentration profiles were fitted consistently using the Monin–Obukhov similarity theory, which are well-known outcomes for stratified turbulent shear flow. The interfacial friction and entrainment coefficients obtained from the fitted profiles agreed well with existing results in literature, confirming the applicability of the Monin–Obukhov theory. Furthermore, a new empirical correlation was proposed for the prediction of a mixing layer thickness.     

Experimental evaluation of the water ingression mechanism for corium cooling

Experiments were performed to assess the significance of water ingression cooling in the quenching of molten corium. Water ingression is a mechanism by which water penetrates into cracks and pores of solidified corium to enhance cooling that would otherwise be severely limited by the low thermal conductivity of the material. Quench tests were conducted with 2100 °C melts weighing 75 kg composed of UO₂, ZrO₂ and chemical constituents of concrete. The amount of concrete in the melts was varied between 4% and 23%. The melts were quenched with an overlying water layer; three tests were conducted at a system pressure of 1 bar and four tests at 4 bar. The measured cooling rates were found to decrease with increasing concrete content and, contrary to expectations, are essentially independent of system pressure. For the lower concrete content melts, cooling rates exceeded the conduction-limited rate with the difference being attributed to the water ingression mechanism. Measurements of the permeability of the corium “ingots” produced by the quench tests were used to obtain a second, independent set of dryout heat flux data, which exhibits the same trend as the quench test data. The data was used to validate an existing dryout heat flux model based on corium permeability associated with thermally induced cracking. The model uses the thermal and mechanical properties of the corium and coolant, and it reproduces the very particular data trend found for the dryout heat flux as a function of concrete content. The model predicts that water ingression cooling would be most effective for concrete-free corium mixtures such as in-vessel type melts. For such a melt the model predicts a dryout heat flux of 400 kW/m² at a pressure of 1 bar. The results of this study provide an experimental basis for a water ingression model that can be incorporated into computer codes used to assess accident management strategies.     

Heat transfer to supercritical pressure carbon dioxide flowing upward through tubes and a narrow annulus passage

A heat transfer test facility, SPHINX, which uses carbon dioxide as a medium at supercritical pressures, has been built at KAERI. A series of experiments are under way for various geometries including tubes of several diameters and narrow annulus passages of a concentric and eccentric layout. The experiments aim to collect heat transfer data and to provide an empirical heat transfer correlation required for a SCWR design. In this paper the test results for tubes of 4.4 mm and 9.0 mm IDs, and a concentric annular passage (8 mm × 10 mm × L1800 mm) are presented for certain combinations of the heat fluxes and mass fluxes. The heat transfer coefficients produced in the tests were compared with those from the existing heat transfer correlations with different media. A new correlation was introduced for the experiment data presented in this paper.     

CFD investigation of helical wire-wrapped 7-pin fuel bundle and the challenges in modeling full scale 217 pin bundle

Flow and temperature distributions of sodium in a heat generating fuel pin bundle with helically wound spacer wire have been predicted from basic principles by solving the three-dimensional conservation equations of mass, momentum and energy, for a wide range of Reynolds number. Turbulence has been modeled using the k– turbulence model. The geometry details of the bundle and heat flux from the fuel pin are similar to that of the Indian Prototype Fast Breeder Reactor (PFBR) that is currently under construction. The focus of the study is to assess the effect of transverse flow in promoting flow and temperature uniformity. It is seen that the ratio of maximum transverse velocity to the maximum axial velocity is nearly equal to the tangent of the rolling up angle of the spacer wire. Due to the wire wrap, the difference in bulk sodium temperature between the peripheral and central sub-channels is reduced to by a factor of 4 when compared to that without spacer wire. The film drop at the junction between wire and the pin is found to be only 70 °C. The predicted results are found to be in close agreement with that of the experimental results reported in literature. The present study considers a 7-pin bundle assembly of one helical pitch. The computational time and memory required for a 217 pin with 15 pitches assembly is ascertained to be 500 times that required for the current study. Hence, research activities have been directed towards developing a parallel CFD code and structural mesh generation software.     

The result of a wall failure in-pile experiment under the EAGLE project

The WF (wall failure) test of the EAGLE program, in which 2 kg of uranium dioxide fuel-pins were melted by nuclear heating, was successfully conducted in the IGR (Impulse Graphite Reactor) of NNC/Kazakhstan. In this test, a 3 mm-thick stainless steel (SS) wall structure was placed between fuel pins and a 10 mm-thick sodium-filled channel (sodium gap). During the transient, fuel pins were heated, which led to the formation of a fuel-steel mixture pool. Under the transient nuclear heating condition, the SS wall was strongly heated by the molten pool, leading to wall failure. The time needed for fuel penetration into the sodium-filled gap was very short (less than 1 s after the pool formation). The result suggests that molten core materials formed in hypothetical LMFBR core disruptive accidents have a certain potential to destroy SS-wall boundaries early in the accident phase, thereby providing fuel escape paths from the core region. The early establishment of such fuel escape paths is regarded as a favorable characteristic in eliminating the possibility of severe re-criticality events. A preliminary interpretation on the WF test results is presented in this paper.