Reactivity feedback coefficients of a material test research reactor fueled with high-density U3Si2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U₃Si₂ dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U₃Si₂ LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δk/k × 10⁻⁵ K⁻¹) −2.116 − 0.118 ρ_U, 0.713 − 37.309/ρ_U and −12.765 − 34.309/ρ_U, respectively for 4.0 ≤ ρ_U (g/cm³) ≤ 6.0.     

The study of reduced transition probabilities for E2 transitions in the decays of Os and Pt nuclei

The ratios of E2 transition rates, i.e., B(E2; I_i − I_f)/B(E2; I_i − I_f) were computed for gamma to ground state band transitions in ¹⁹²Os and ¹⁹²Pt. The reduced transition probabilities for the E2 transitions evaluated on the basis of Interacting Boson Approximation (IBA-I) model and those measured experimentally are compared. The SU(3) symmetry of IBA-I is not strictly obeyed by the nuclei in which T(E2; I_i − I_f) transitions are observed, i.e., the symmetry SU(3) is broken. The percentage sum coincidence corrections are applied to marginalise contributions from crossover transitions and the intensities of affected transitions are found to agree fairly well with earlier sum peak method applications to the decay of ¹⁹²Ir.     

Propagation of uncertainties in problems of structural reliability

The problem of controlling a variable Y such that the probability of its exceeding a specified design limit L is very small, is treated. This variable is related to a set of random variables X_i by means of a known function Y = ƒ(X_i). The following approximate methods are considered for estimating the propagation of error in the X_i's through the function ƒ(·): linearization; method of moments; Monte Carlo methods; numerical integration. Response surface and associated design of experiments problems as well as statistical inference problems are discussed.     

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.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

Single and multiple discharge from a stratified two-phase region through small branches

Experimental data are presented for the mass flow rate and quality during single, dual and triple discharge from a stratified air–water region through small side branches (d=6.35 mm) installed on a semicircular wall. Dimensions of the semicircular wall and branches were chosen such that interaction among the branches is possible under certain flow conditions. All the branches were adjusted to have the same hydraulic resistance (R=1000 (kg m)^−1/2) and for the cases of dual and triple discharge, the same pressure drop ΔP was imposed across all active branches. Tests were conducted at two system pressures P₀=316 and 517 kPa and the pressure drop was varied within the range 40≤ΔP≤235 kPa. Data analysis is presented with emphasis on the effect of wall curvature and also the effect of additional discharges on the flow from a certain branch. The present data can serve as benchmark data for testing numerical safety codes and they should guide future research on the flow from two-phase headers.     

JR curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents in oxygen-containing high temperature water at 240°C

J_R curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents (0.003 and 0.011 wt.%) were determined at 240°C in oxygen-containing high temperature water as well as in air. The tests were performed by the single-specimen unloading compliance technique at load line displacement rates from 1 × 10⁻⁴ down to 1 × 10⁻⁶ mm s⁻¹ on 20% side-grooved 2T CT specimens in an autoclave testing facility at an oxygen content of 8 ppm and a pressure of 7 MPa under quasi-stagnant flow conditions.In the case of testing in high temperature water, remarkably lower J_R curves than in air at the same load line displacement rate (1 × 10⁻⁴ mm s⁻¹) were obtained. A decrease in the load line displacement rate as well as an increase in the sulphur content of the steel caused a reduction of the J_R curves. At the fastest load line displacement rate a stretch zone could be detected fractographically on the specimens tested in air and in high temperature water and consequently J_i could be determined. When testing in high temperature water, the J_i value of the higher sulphur material type decreases from 45 N mm⁻¹ in air to 3 N mm⁻¹, much more than that of the optimized material type from 51 N mm⁻¹ in air to 20 N mm⁻¹ at 1 × 10⁻⁴ mm s⁻¹.     

Turbulent heat mixing of a heavy liquid metal flow in the MEGAPIE target geometry—The heated jet experiment

The MEGAPIE target installed at the Paul–Scherrer Institute is an example of a spallation target using eutectic liquid lead–bismuth (Pb⁴⁵Bi⁵⁵) both as coolant and neutron source. An adequate cooling of the target requires a conditioning of the flow, which is realized by a main flow transported in an annular gap downwards, u-turned at a hemispherical shell into a cylindrical riser tube. In order to avoid a stagnation point close to the lowest part of the shell a jet flow is superimposed to the main flow, which is directed towards to the stagnation point and flows tangentially along the shell.The heated jet experiment conducted in the THEADES loop of the KALLA laboratory is nearly 1:1 representation of the lower part of the MEGAPIE target. It is aimed to study the cooling capability of this specific geometry in dependence on the flow rate ratio (Q_main/Q_jet) of the main flow (Q_main) to the jet flow (Q_jet). Here, a heated jet is injected into a cold main flow at MEGAPIE relevant flow rate ratios. The liquid metal experiment is accompanied by a water experiment in almost the same geometry to study the momentum field as well as a three-dimensional turbulent numerical fluid dynamic simulation (CFD). Besides a detailed study of the envisaged nominal operation of the MEGAPIE target with Q_main/Q_jet = 15 deviations from this mode are investigated in the range from 7.5 ≤ Q_main/Q_jet ≤ 20 in order to give an estimate on the safe operational threshold of the target.The experiment shows that, the flow pattern establishing in this specific design and the turbulence intensity distribution essentially depends on the flow rate ratio (Q_main/Q_jet). All Q_main/Q_jet-ratios investigated exhibit an unstable time dependent behavior. The MEGAPIE design is highly sensitive against changes of this ratio.Mainly three completely different flow patterns were identified. A sufficient cooling of the lower target shell, however, is only ensured if Q_main/Q_jet ≤ 12.5. In this case the jet flow covers the whole lower shell. Although for Q_main/Q_jet ≤ 12.5 the flow is more unstable compared to the other patterns most of the fluctuations close to the centerline are in the high frequency range (>1 Hz), so that they will not lead to severe temperature fluctuations in the lower shell material. In this case the thermal mixing occurs on large scales and is excellent.For flow rate ratios Q_main/Q_jet > 12.5 complex flow patterns consisting of several fluid streaks and vortices were identified. Since in these cases the jet flow does not fully cover the lower shell an adequate cooling of the MEGAPIE target cannot be guaranteed and thus temperatures may appear exceeding material acceptable limits.All conducted experiments show a high sensitivity to asymmetries even far upstream. A comparison of the numerical simulation, which assumed a symmetric flow, with the experimental data was due to the experimentally found asymmetry only partially possible.     

Simulations of turbulence, heat transfer and mixing across narrow gaps between rod-bundle subchannels

Turbulent flow and temperature fields were determined numerically in a rectangular duct containing a heated rod. As the spacing δ between the rod and the duct wall decreased from 0.10D (D is the rod diameter) to 0.03D, coherent turbulent kinetic energy and temperature fluctuations dramatically increased in the gap region, but, for δ = 0.01D, coherent fluctuations essentially disappeared. As δ/D → 0, the frequency of coherent fluctuations decreased and cross-gap mixing weakened, contrary to predictions based on extrapolated available empirical correlations.     

Computation of Mass Sweeping Efficiency Factor and Current Efficiency Factor in Z-pinch Devices

In this paper, mass sweeping efficiency factor (f _m ) and current efficiency factor (f _i ) have been computed for Z-pinch devices. We used slug model for analysis of Z-pinch dynamics. Magnetic piston reaps electrons and ions in duration of motion. But only a fraction of plasma mass sweeps with magnetic piston, therefore we should add mass sweeping efficiency factor (f _m ) in equations. Such like alone the fraction of electrical current flows of magnetic piston and remainder of it flows of internal and external radial of magnetic piston, so we should add f _i in equations. In this paper, equations are solved with characteristics of CERN Z-pinch device (its length and radius, resistivity, circuit inductance and capacitanc and plasma inductance) and with values of Boggasch experiments (discharge voltage: 15 kV, initial pressure: 400 pa). Recorded code runs with different values of f _m and f _i and in each section, pinch time and pinch current are compared with Boggasch experimental values. Optimum values for f _m and f _i obtain with Comparing between numerical values and experimental values. These values are f _i  = 0.8 and f _m  = 0.08.     

Comparison of experimental and finite element analytical results for the strength and the deformation of pipes with local wall thinning subjected to bending moment

Fracture behaviors of pipes with local wall thinning are very important for the integrity of power plant piping system. In this study, monotonic bending tests without internal pressure are conducted on 48.6 mm diameter Schedule 80 (thickness, 5.1 mm) STS370 full-scale carbon steel pipe specimens. Fracture strengths of locally wall-thinned pipes were calculated by elasto-plastic analysis using finite element method. The elasto-plastic analysis was performed by FE code ANSYS. We simulated various types of local wall thinning that can be occurred at pipe surface due to coolant flow. Locally wall thinned shapes were machined to be different in size along the circumferential or axial direction of straight pipes. We investigated fracture strengths and failure modes of locally wall thinned pipes by four point bending test. And, the allowable limit of pipes with local wall thinning was investigated. In addition, we compared the simulated results by finite element analysis with experimental data. The failure mode, fracture strength and fracture behavior obtained from FE analyses showed well agreement with experimental results. From the test results, we identified three types of failure modes into ovalization, local buckling and crack initiation. These failure modes could be classified according to thinned depth, thinned length and thinned angle of a pipe. For locally wall-thinned specimens, maximum moments (M_max) were estimated by using the net-section stress criterion. Pipes with local wall thinning can be estimated using σ_u instead of σ_f because of 1.19σ_f  σ_u. Also, the axial strain affects failure modes occurred on local wall thinning. the allowable limit of local wall thinning for carbon steel pipe used can be given as follows; in the case of M_max ≥ M_y, if 10 ≤ l < 25 mm, d/t can be allowed to about 55%, and if 25 ≤ l < 100 mm, d/t can be allowed to about 50%. Also, if 100 ≤ l ≤ 120 mm, d/t can be allowed to about 29%.     

Prospects for fission gas extraction from in-reactor UO2 nuclear fuel elements

Due to the many problems encountered in the design of fuel rods for the safe operation of commercial nuclear reactors, caused by the fission gases generated by the fission of fissile material, it was considered opportune to make a theoretical analysis of the feasibility of extraction of fission gases from the fuel rod while in operation.This analysis in the steady state of a Zircaloy-2 sheathed fuel rod containing UO₂ as a fuel, with a 2 mm (2.7 vol.%) diameter porous graphite cylinder inserted in the centre, has demonstrated that a total volume of fission gases (xenon, krypton, and iodine) of about 1.1 × 10⁻⁶ cm³/s (at STP) can be extracted from the fuel rod at a controlled rate, determined by the inherent property of fission gas migration towards the centre of the fuel rod from its place of formation. In this analysis, the fuel rod was assumed to be subjected to irradiation in a reactor the size of a Bruce “A” reactor, operating at 3000 megawatts thermal power. The extracted volume of gas was calculated on a 900 h cycle after the first 90 h of reactor operation had elapsed.     

Some problems for bundle CHF prediction based on CHF measurements in simple flow geometries

A comparison of critical heat flux (CHF) fuel bundles data with CHF data obtained in simple flow geometries was made. The base for the comparison was primary experimental data obtained in annular, circular, rectangular, triangular, and dumb-bell shaped channels cooled with water and R-134a. The investigated range of flow parameters (pressure, mass flux, and critical quality) in R-134a was chosen to be equivalent to modern nuclear reactor water flow conditions (p=7 and 10 MPa, G=350–5000 kg (m² s)⁻¹, x_cr=−0.1–1). The proper scaling laws were applied to convert the data from water to R-134a equivalent conditions and vise versa. The effects of flow parameters (p, G, x_cr) and the effects of geometric parameters (D, L) were evaluated during comparison. The comparison showed that no one simple flow geometry can be used for accurate and reliable bundle CHF prediction in wide range of flow parameters based on local (critical) conditions approach. The comparison also showed that the limiting critical quality phenomenon is unique characteristic for each flow geometry which depends on many factors: flow conditions (pressure and mass flux), geometrical parameters (diameter or surface curvature, gap size, etc.), flow obstructions (spacers, appendages, turbulizers, etc.) and others.     

Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters in (D–T) driven hybrid reactor system

Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters are investigated for different coolants under various first wall loads (P_w=2, 5, 7, 8, 9, and 10 MW m⁻²) in (D, T) (deuterium and tritium) driven and fueled with UO₂ hybrid reactors. Plasma chamber dimension, DR, with a line fusion neutron source is 300 cm. The fissile fuel zone is considered to be cooled with four different coolants with various volume fractions, the volumetric ratio of coolant-to-fuel [(Vm/V_f) = 1:2, 1:1, and 2:1], gas (He, CO₂), flibe (Li₂BeF₄), natural lithium (Li), and eutectic lithium (Li₁₇Pb₈₃). Calculation in the fuel rods and the behavior of the fissile fuel have been observed during 4 years for discrete time intervals of Δt=15 days and by a plant factor (PF) of 75%. As a result of the calculation, cumulative fissile fuel enrichment (CFFE) value indicating rejuvenation performance has increased by increasing P_w for all coolants and . Although CFFE and neutronic performance parameter values increase to the higher values by increasing P_w, the maximum temperature in the centerline of the fuel roads has exceeded the melting point (T_m>2830°C) of the fuel material during the operation periods. However, the best CFFE (11.154%) is obtained in gas coolant blanket for =1:2 (29.462% coolant, 58.924% fuel, 11.614% clad), under 10 MW m⁻² first wall load, followed by flibe with CFFE=11.081% for =2:1 (62.557% coolant, 31.278% fuel, 6.165% clad), under 7 MW m⁻², and flibe with CFFE=9.995% for =1:1 (45.515% coolant, 45.515% fuel, 8.971% clad), under 7 MW m⁻² during operation period without reaching the melting point of the fuel material. While maximum CFFE value has been obtained in fuel rod row#10 in gas, natural lithium, and eutectic lithium coolant blankets, it has been obtained in fuel rod row#1 in flibe coolant blanket for all and P_w. At the same condition, the best neutronic performance parameter values, tritium breeding ratio (TBR)= 1.4454, energy multiplication factor (M)= 9.2018, and neutron leakage (L)= 0.0872, have been obtained in eutectic lithium coolant blankets for the =1:2, followed by gas, natural lithium, and flibe coolant blankets. The isotopic percentage of ²⁴⁰Pu is higher than 5% in all blankets for P_w 7 MW m⁻², so that plutonium component in all blankets can be never reach a nuclear weapon grade quality during the operation period.     

CHF and dryout point in vertical narrow annuli

Experimental study associated with CHF and dryout point in narrow annuli is conducted with 1.5 mm and 1.0 mm gap, respectively. Distilled water is used as work fluid. The parameters examined were: pressure from 2.0 MPa to 4.0 MPa; mass flux from 26.0 kg/(m² s) to 69.0 kg/(m² s); heat flux from 10 kW/m² to 70 kW/m²; exit equilibrium mass quality from 0.52 to 1.08.It is found that CHF monotonously increases with mass flux in internally heated annuli and bilaterally heated annuli. However, the observed trends are not similar to that in externally heated annuli. The CHF is not affected significantly by mass flux.Critical qualities of dryout point (X_DO) decreases with mass flux and increases with inlet qualities. Under the same conditions X_DO in outer tube are always larger than that in inner tube. According to experimental data, a criterion for the appearance of dryout point for bilaterally heated has been presented.The comparison with the correlations [КУТАТЕЛАДЗЕ, C.C., 1979. Тедплоэнергетика, No. 6] and experimental data indicates that the existing correlations applied to tube cannot predict X_DO in narrow annuli well. Based on experimental data, a new correlation is developed.     

Ion irradiation induced nanocrystal formation in amorphous Zr55Cu30Al10Ni5 alloy

Ion irradiation can be used to induce partial crystallization in metallic glasses to improve their surface properties. We investigated the microstructural changes in ribbon Zr₅₅Cu₃₀Al₁₀Ni₅ metallic glass after 1 MeV Cu-ion irradiation at room temperature, to a fluence of 1.0 × 10¹⁶ cm⁻². In contrast to a recent report by others that there was no irradiation induced crystallization in the same alloy [S. Nagata, S. Higashi, B. Tsuchiya, K. Toh, T. Shikama, K. Takahiro, K. Ozaki, K. Kawatusra, S. Yamamoto, A. Inouye, Nucl. Instr. and Meth. B 257 (2007) 420], we have observed nanocrystals in the as-irradiated samples. Two groups of nanocrystals, one with diameters of 5–10 nm and another with diameters of 50–100 nm are observed by using high resolution transmission electron microscopy. Experimentally measured planar spacings (d-values) agree with the expectations for Cu₁₀Zr₇, NiZr₂ and CuZr₂ phases. We further discussed the possibility to form a substitutional intermetallic (Ni_xCu_1−x)Zr₂ phase.     

A bulk tungsten divertor row for the outer strike point in JET

In the frame of the ITER-like wall project, a new row of divertor tiles has been developed which consists of 96 bulk tungsten load-bearing septum replacement plates (LB-SRP). Exposed to the outer strike point for most ITER-relevant, high triangularity configurations, they shall be subject to high power loads (locally 10 MW/m² and above). These conditions are demanding, particularly for an inertially cooled design as prescribed. The expected erosion rates are high as well as the risk of melting, especially with transients and repetitive ELM loads. The development is also a real challenge with respect to the inevitable excursions of the tungsten material through the so-called DBTT, ductile-to-brittle transition temperature.A lamella design has been selected to fulfil the requirements with respect to the thermo-mechanical and electromagnetic loads during disruptions (∂T/∂z ≤ 5 × 10⁴ K/m vertically, induction rate of change ∂B/∂t ≤ 100 T/s, and I_halo ≤ 18 kA/module). Care is taken to act on refractory metals solely with compressive forces to a large extent. The dedicated clamping concept is described. Results of a test exposure to an electron beam around 70 MJ/m² substantiate the resort to ‘high temperature’ materials like – among others – high-grade Nimonic^® alloys, molybdenum or ceramic coatings.     

Transformation and fragmentation behavior of molten metal drop in sodium pool

In order to clarify the fragmentation mechanism of a metallic alloy (U–Pu–Zr) fuel on liquid phase formed by metallurgical reactions (liquefaction temperature = 650 °C), which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum (melting point = 660 °C) and sodium mainly under the condition that the boiling of sodium does not occur. When the instantaneous contact interface temperature (T_i) between molten aluminum drop and sodium is lower than the boiling point of sodium (T_c,bp), the molten aluminum drop can be fragmented and the mass median diameter (D_m) of aluminum fragments becomes small with increasing T_i. When T_i is roughly equivalent to or higher than T_c,bp, the fragmentation of aluminum drop is promoted by thermal interaction caused by the boiling of sodium on the surface of the drop. Furthermore, even under the condition that the boiling of sodium does not occur and the solid crust is formed on the surface of the drop, it is confirmed from an analytical evaluation that the thermal fragmentation of molten aluminum drop with solid crust has a potential to be caused by the transient pressurization within the melt confined by the crust. These results indicate the possibility that the metallic alloy fuel on liquid phase formed by the metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt.     

Fluid-to-fluid modeling of critical heat flux in 4 × 4 rod bundles

Fluid-to-fluid modeling of critical heat flux (CHF) is to simulate the CHF behaviors for water by employing low cost modeling fluid, and the flow scaling factor is the key to apply the technique to fuel bundles. The CHF experiments in 4×4 rod bundles have been carried out in Freon-12 loop in equivalent nuclear reactor water conditions (P=10.0–16.0 MPa, G=488.0–2100.0 kg/m² s, X_cr=−0.20–0.30). The models in fluid-to-fluid modeling of CHF is verified by the CHF data for Freon-12 obtained in the experiment and the CHF correlation for water obtained by Nuclear Power Institute of China (NPIC) in the same 4×4 rod bundles. It has been found that the S.Y. Ahmad Compensation Distortion model, the Lu Zhongqi model, the Groeneveld model and Stevens–Kirby model overpredict the bundles CHF values for water. Then an empirical correlation of flow scaling factor is proposed. Comparison of the CHF data in two kinds of test sections for Freon-12, in which the distance of the last grid away the end of heated length is different, shows that the spacer grid, which is located at 20 mm away from the end of the heated length, has evidently influenced on the CHF value in the 4×4 rod bundles for Freon-12. This is different from that for water, and the need for further work is required.