المؤلفون: Alis Musa, Guido Mazzini, Rostilav Fukac

المصدر: Journal of Nuclear Engineering and Radiation Science. 7

مصطلحات موضوعية: Thermal efficiency, Radiation, 020209 energy, Nuclear engineering, Heat losses, 02 engineering and technology, 010403 inorganic & nuclear chemistry, 01 natural sciences, 0104 chemical sciences, Thermal hydraulics, Loop (topology), Nuclear Energy and Engineering, Heat transfer, 0202 electrical engineering, electronic engineering, information engineering, Code (cryptography), Environmental science, Leakage (electronics)

الوصف: Supercritical water (SCW) has advantages like high thermal efficiency and can operate at high temperature and pressure. At the same time, however, these properties bring up related issues, such as material compatibility and corrosion resistance. In an effort to fully investigate the operating conditions, and solutions to these issues, test facilities are being built by many research organizations. One such organization, the Research Center Řež (CVR) located in the Czech Republic, has developed an experimental supercritical water loop (SCWL). The purpose of this loop is to provide experimental data from material testing in various conditions, including operating under the neutron field. This will be achieved by inserting a test channel into the existing experimental reactor light water reactor 15 (LVR-15), which will require a license from the state nuclear regulator (State Office for Nuclear Safety (SUJB)). Part of the licensing documentation is the safety analysis, which combines results from developed models using the thermohydraulic code ATHLET 3.1 A patch 1, as well as the experimental out of pile data. Among the postulated scenarios, an abnormal sequence (labeled A2—Loss of power in the loop) was analyzed in order to provide a preliminary benchmark. This scenario is similar to the postulated in-pile A2 and it was used for the benchmark activity. The aim of this paper is to present this activity including the adopted assumptions in the model. In particular, the paper presents, how these assumptions influenced the results indicating the discrepancies obtained in the first part of the transient.

URL الوصول: https://explore.openaire.eu/search/publication?articleId=doi_________::76a10af7fdc876e858514e1aa74ab694
https://doi.org/10.1115/1.4050207

المؤلفون: Béla Hegyesi, Attila Kiss, András Szabolcs Ványi, Patrik Richárd Ujváry, Gyula Csom

المصدر: Journal of Nuclear Engineering and Radiation Science. 6

مصطلحات موضوعية: Radiation, Materials science, business.industry, 020209 energy, Nuclear engineering, Thorium, chemistry.chemical_element, 02 engineering and technology, Computational fluid dynamics, 010403 inorganic & nuclear chemistry, 01 natural sciences, 0104 chemical sciences, Thermal hydraulics, Nuclear Energy and Engineering, chemistry, 0202 electrical engineering, electronic engineering, information engineering, Boundary value problem, business

الوصف: Inherently poorer moderation in supercritical water-cooled reactors (SCWRs) due to average density lower than in light water reactors and the resulted spectral shift can be useful when we apply thorium fuel-cycle instead of uranium–plutonium one, according to an ongoing study in Budapest University of Technology and Economics (BME) Institute of Nuclear Techniques (NTI). Upon this conclusion, a thorium-fueled SCWR design (Th-SCWR) has been proposed by BME NTI. In the current feasibility study phase, detailed three-dimensional (3D) computational fluid dynamics (CFD) calculations with novel neutronics analysis were coupled and conducted separately. Neutronics calculations provided the distribution of heat source, while the CFD analysis gave back axial distribution of coolant density (this iteration was repeated until an acceptable convergence). This paper presents the CFD analysis on thermal hydraulics of the initial design (two CFD models without any spacer device and one model with wrapped wire spacer) of Th-SCWR fuel assembly. As results of the preliminary design of Th-SCWR cladding wall, coolant and fuel temperatures have been determined; the flow field with and without spacer device has been showed, and the application of wrapped wire spacer has been proposed.

URL الوصول: https://explore.openaire.eu/search/publication?articleId=doi_________::a439fc690f8332c455c63698fe4534f0
https://doi.org/10.1115/1.4046904

المؤلفون: Bence Mervay, Attila Kiss

المصدر: Journal of Nuclear Engineering and Radiation Science. 6

مصطلحات موضوعية: Thermal hydraulics, Radiation, Materials science, Nuclear Energy and Engineering, 020209 energy, Numerical analysis, Bundle, 0202 electrical engineering, electronic engineering, information engineering, 02 engineering and technology, Mechanics, 010403 inorganic & nuclear chemistry, 01 natural sciences, 0104 chemical sciences

الوصف: The application of relatively simple and cheap wrapped wire spacer in the European supercritical water-cooled reactor (SCWR) (high-performance light water reactor (HPLWR)) has been proposed in order to provide enhanced heat transfer in the fuel assembly without unacceptable penalty in pressure loss. The wires cause twisting flow in the fuel assembly, which means the coolant not only flows straight in the axial direction but also has a significant transverse velocity component, and strong mixing between neighboring subchannels occurs. The aim of this ongoing research is to numerically investigate the effect of wrapped wire spacers on thermal hydraulics of the turbulent coolant flow and its heat transfer in a small bundle of four fuel rods. One bare and six-wired geometries with varying wire pitches (1–6 turn(s) of wires) have been studied. It was found that the wires generate significant amount of transverse velocity, decrease the wall temperature, and increase the heat transfer coefficient mostly in corner subchannels which were the hottest in bare geometry. Thus, the presence of wires enhances heat transfer where it is most needed. Temperature hot spots with moderate values have been identified on the cladding wall of fuel rods. Based on the results, a technically optimal choice of number of wire turns from thermal hydraulic sense has been proposed.

URL الوصول: https://explore.openaire.eu/search/publication?articleId=doi_________::44cf94582f29870b95d6b7491bf9dcf7
https://doi.org/10.1115/1.4046842

المؤلفون: Obaidurrahman K., Vivek A. Kale

المصدر: Journal of Nuclear Engineering and Radiation Science. 6

مصطلحات موضوعية: Physics, Radiation, 020209 energy, Nuclear engineering, Dynamics (mechanics), 02 engineering and technology, 010403 inorganic & nuclear chemistry, 01 natural sciences, 0104 chemical sciences, Coolant, Thermal hydraulics, Nuclear Energy and Engineering, 0202 electrical engineering, electronic engineering, information engineering, Code (cryptography), Neutron, Transient (computer programming), Reactivity (chemistry)

الوصف: This paper presents the analysis of reactivity initiated transients in an idealized, light water research reactor as a part of International Atomic Energy Agency (IAEA) safety related benchmark. The simulation model is based on point reactor kinetics coupled with one-dimensional (1D), two-channel model for thermal hydraulics. The point kinetics equations (PKEs) have been solved using an implicit Runge–Kutta (RK) method and the coolant transport equations have been solved using implicit finite difference formulation. Accuracy of the implemented models and methods has been demonstrated. Important safety parameters like peak power, peak fuel, and coolant temperatures have been predicted for a series of transients. Intercode comparison shows that the predictions of the present simulations are in good agreement with other codes. This approach provides a time efficient solution for safety analysis of reactors with tightly coupled core where point kinetics can be applied. To address the sensitivity of predictions with respect to important input parameters, simulations have been carried out with different sets of inputs reported in the literature. They indicate that predictions for fast transients are spread over a wider range compared to slow transients. For a given transient, predictions of peak power have a wider spread, while peak temperatures are relatively less sensitive to neutronic inputs. Also, for fast transients, prompt neutron generation time and delayed neutron fraction have dominant influence on the evolution of power. For slow transients, the reactivity feedback effects are equally important.

URL الوصول: https://explore.openaire.eu/search/publication?articleId=doi_________::8b75648ef37cd7d2c97e7c713e64680d
https://doi.org/10.1115/1.4045394

المؤلفون: David R. Novog, David William Hummel

المصدر: Journal of Nuclear Engineering and Radiation Science. 5

مصطلحات موضوعية: Radiation, Materials science, 020209 energy, Nuclear engineering, Shutdown, 02 engineering and technology, 010403 inorganic & nuclear chemistry, 01 natural sciences, 0104 chemical sciences, Coolant, Thermal hydraulics, Nuclear Energy and Engineering, 0202 electrical engineering, electronic engineering, information engineering, Neutron, Neutron kinetics

الوصف: The Canadian supercritical water-cooled reactor concept features a re-entrant fuel channel wherein coolant first travels down a center flow tube and then up around the fuel elements. Previous work demonstrated that in cases of sudden coolant flow reduction or reversal (such as that which would result from a large pipe break near the core inlet), the coolant density reduction around the fuel has a positive reactivity effect that results in a power excursion. Such a transient is inherently self-terminating since the inevitable density reduction in the center flow tube has a very large negative reactivity effect. Nevertheless, a brief power pulse would ensue. In this work, the possibility of mitigating the power pulse with a fast-acting shutdown system was explored. The shutdown system model, consisting of bottom-inserted neutron absorbing blades and realistic estimates of insertion rates and trip conditions, was added to a full-core coupled spatial neutron kinetics and thermal-hydraulics model. It was demonstrated that such a system can effectively mitigate both the peak magnitude of the power excursion and its duration.

URL الوصول: https://explore.openaire.eu/search/publication?articleId=doi_________::ba88b7ea8186dc09e386ad064ab15f98
https://doi.org/10.1115/1.4042597

المؤلفون: Hiroyuki Ohshima, Norihiro Kikuchi, Norihiro Doda, Yasutomo Imai, Ryuji Yoshikawa, Masaaki Tanaka

المصدر: Journal of Nuclear Engineering and Radiation Science. 5

مصطلحات موضوعية: Pressure drop, Radiation, Materials science, Buoyancy, Turbulence, 020209 energy, Sodium, chemistry.chemical_element, 02 engineering and technology, Mechanics, engineering.material, 010403 inorganic & nuclear chemistry, 01 natural sciences, 0104 chemical sciences, Thermal hydraulics, Sodium-cooled fast reactor, Nuclear Energy and Engineering, chemistry, 0202 electrical engineering, electronic engineering, information engineering, engineering, Duct (flow)

الوصف: In the design study of advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) called FA with inner duct structure (FAIDUS) is expected to enhance reactor safety during a core-disruptive accident. Evaluating the thermal-hydraulics in FAIDUS under various operating conditions is required for its design. This study is the first step toward confirming the design feasibility of FAIDUS; the thermal-hydraulics in FAIDUS are investigated with an in-house subchannel analysis code called asymmetrical flow in reactor elements (ASFRE), which can be applied to a wire-wrapped fuel pin bundle in conjunction with the distributed resistance model (DRM) and the turbulence-mixing model of the Todreas–Turi correlation model (TTM). Before simulating the thermal-hydraulics in FAIDUS, a few validations of DRM and TTM are conducted by comparing the numerical results of the pressure drop coefficients or temperature distribution obtained using ASFRE with the experimental data obtained using an apparatus with water or sodium for simulated FAs. After these validations, thermal-hydraulic analyses of FAIDUS and a typical FA are conducted for comparison. The numerical results indicate that, at a high flow rate simulating rated operation condition, no significant asymmetric temperature and velocity distribution occur in FAIDUS compared to the distribution in the typical FA. In addition, at a low flow rate simulating natural circulation condition in decay heat removal, the temperature distribution in FAIDUS is similar to that in the typical FA. This is because the local flow acceleration and the flow redistribution due to buoyancy force may occur in FAIDUS and the typical FA.

URL الوصول: https://explore.openaire.eu/search/publication?articleId=doi_________::970474ca37d3f4a7c3952ca145298f49
https://doi.org/10.1115/1.4042191

المؤلفون: Patricio Alberto, Alicia Doval, Jorgelina Lupiano Contreras

المصدر: Journal of Nuclear Engineering and Radiation Science. 5

مصطلحات موضوعية: Thermal hydraulics, Radiation, Materials science, Nuclear Energy and Engineering, Flow (mathematics), 020209 energy, 0202 electrical engineering, electronic engineering, information engineering, 02 engineering and technology, Mechanics, Heat transfer coefficient, 010403 inorganic & nuclear chemistry, 01 natural sciences, 0104 chemical sciences

الوصف: Difficulties are experienced during the thermal–hydraulic design of a nuclear reactor operating in the transition flow regime and are the result of the inability to accurately predict the heat transfer coefficient (HTC). Experimental values for the HTC in rectangular channels are compared with the calculated by correlations usually used for the design of material testing reactors (MTR). The values predicted by Gnielinski and Kreith correlations at Reynolds numbers below 5000 are not necessarily conservative. The Al-Arabi-Churchill correlation with the correction proposed by Jones has proved to be conservative for Reynolds between 2100 and 5000. Two alternative design approaches are proposed to solve a specific thermal–hydraulic design problem for a MTR operating at Reynolds 2500. The conservative approach comprises two alternatives: the use of Al-Arabi correlation with no uncertainty factors, as it has proved to be conservative, or the use of Kreith correlation with a maximum uncertainty. In this conservative approach, maximum deviations in other input parameters are also taken into account. The best estimate plus uncertainty approach considers an uncertainty distribution in input parameters to generate a random sample of 59 inputs. An uncertainty distribution based on the ratio between the experimental and the calculated HTC, when using Kreith correlation, is considered. Results are given in terms of maximum and minimum bounds for the figure of merit used as design criterion with 95% probability and 95% confidence level. The best estimate plus uncertainty approach offers a less penalizing design and its use depends on regulator's acceptance.

URL الوصول: https://explore.openaire.eu/search/publication?articleId=doi_________::74f60328ef7e37b8ed3a5253624571af
https://doi.org/10.1115/1.4042363