Boiling Transition Research Articles

Experimental investigation and quasi-steady modeling of nucleate boiling in mini-channel thermosyphons

This study explores the understanding of two-phase cooling in thermosyphons using a quasi-steady state methodology during the boiling process. A thermosyphon is built as a passive loop to dissipate heat with an evaporator and a condenser. The evaporator is made of multiple mini channels (hydraulic diameter of 1.54 mm, length of 260 mm with 7 × 2 internal ports for a Confinement number of 0.55) with HFO-1336mzz(E) as the working fluid. Different heat loads (500 – 4000 W) are generated directly on the external contact surface of the evaporator to create all the different stages of boiling, from monophasic regime to steady nucleate boiling with the Onset of Nucleate Boiling transition. The temperature evolution inside the evaporator is measured at different heights and compared with a theoretical assumption of a quasi-steady state. A characteristic time depending on critical factors such as thermal mass is determined to model the temperature during a generated heat load. A good agreement between experimental measurements and the quasi-steady model is shown. Thus, this study emphasizes tracking the temperature evolution over time within the system. This dynamic perspective offers a nuanced understanding of the system’s response to varying heat inputs, transient phases, and the onset of boiling. This characterized local behavior provides an original insight into the boiling appearing inside mini-channels. It is shown that boiling is initiated by nucleation at a few specific sites and then propagates up to a critical length due to high vapor generation, introducing a thermal lag during the boiling incipience.

Assessment of CTF Boiling Transition and Critical Heat Flux Modeling Capabilities Using the OECD/NRC BFBT and PSBT Benchmark Databases

Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.

Thermal Feasibility Analyses for the 1356MWe High Conversion-Type Innovative Water Reactor for Flexible Fuel Cycle

A new reactor concept of innovative water reactor for flexible fuel cycle (FLWR) is under development at Japan Atomic Energy Agency in cooperation with Japanese reactor suppliers. A design of 1,356 MWe high conversion boiling water reactor-type FLWR core, which has an instantaneous conversion ratio of 1.04, negative void coefficient, high burnup of 65 GWd/t, and 15-month operational cycle length, has been constructed. So far, studies on thermal-hydraulic characteristics have been performed for tight lattice core. Evaluation methods for the critical power and the pressure drop under both the steady and the transient states have been established, and a modified TRAC-BF1 code has been developed for the thermal-hydraulic design of the FLWR. In this paper, the thermal feasibility of the designed 1356MWe FLWR core is analyzed by using the modified TRAC-BF1 code. The analysis is first carried out for the current core design. It is confirmed that no boiling transition (BT) occurs under the steady state. However, the minimum critical power ratio (MCPR) is only about 1.08, and the BT is confirmed occurring under the postulated abnormal transient processes. Therefore, concretizations of the conditions that ensure the thermal feasibility of a natural circulation-type FLWR and a forced circulation-type FLWR are performed. As for the results, for a forced circulation-type FLWR, the operation-limited MCPR (OLMCPR) is 1.32, and the necessary minimum core coolant flow rate is 640 kg/(m2s). For a natural circulation-type FLWR, the OLMCPR is 1.19, and the necessary minimum core coolant flow rate is 560 kg/(m2s).

Effect of Rod Bowing on Critical Power Based on Tight-Lattice 37-Rod Bundle Experiments

The confirmation of thermal-hydraulic performance is one of the most important R&D requirements for the design of the Innovative Water Reactor for FLexible Fuel Cycle (FLWR). Since the effect of rod bowing on critical power has not been determined yet due to the lack of experimental data, a large-scale thermal-hydraulic experiment using a tight-lattice 37-rod bundle test section with a bowed rod was carried out with pressure ranging from 2–9 MPa and mass velocity at 200–1000 kg/(m2s). It was confirmed that boiling transition (BT) occurs downstream of the rod contact point, and that the wall temperature trace during the BT follows the typical BT pattern of BWR. The critical power with a bowed rod is about 10% lower than that without rod bowing. The critical power increases monotonically with the increase in mass velocity, with the decrease in inlet water temperature, and with the decrease in exit pressure, and these trends are similar to those of the basic bundle without rod bowing. Thus, there is a negligible effect of rod bowing on the dependence of critical power on the mass velocity, the inlet temperature, and the exit pressure.

Effect of Rod Bowing on Critical Power Based on Tight-Lattice 37-Rod Bundle Experiments

The confirmation of thermal-hydraulic performance is one of the most important R&D requirements for the design of the Innovative Water Reactor for FLexible Fuel Cycle (FLWR). Since the effect of rod bowing on critical power has not been determined yet due to the lack of experimental data, a large-scale thermal-hydraulic experiment using a tight-lattice 37-rod bundle test section with a bowed rod was carried out with pressure ranging from 2–9 MPa and mass velocity at 200–1000 kg/(m2s). It was confirmed that boiling transition (BT) occurs downstream of the rod contact point, and that the wall temperature trace during the BT follows the typical BT pattern of BWR. The critical power with a bowed rod is about 10% lower than that without rod bowing. The critical power increases monotonically with the increase in mass velocity, with the decrease in inlet water temperature, and with the decrease in exit pressure, and these trends are similar to those of the basic bundle without rod bowing. Thus, there is a negligible effect of rod bowing on the dependence of critical power on the mass velocity, the inlet temperature, and the exit pressure.

Numerical Investigation of Cross Flow Phenomena in a Tight-Lattice Rod Bundle Using Advanced Interface Tracking Method

In relation to the design of an innovative FLexible-fuel-cycle Water Reactor (FLWR), investigation of thermal-hydraulic performance in tight-lattice rod bundles of the FLWR is being carried out at Japan Atomic Energy Agency (JAEA). The FLWR core adopts a tight triangular lattice arrangement with about 1 mm gap clearance between adjacent fuel rods. In view of importance of accurate prediction of cross flow between subchannels in the evaluation of the boiling transition (BT) in the FLWR core, this study presents a statistical evaluation of numerical simulation results obtained by a detailed two-phase flow simulation code, TPFIT, which employs an advanced interface tracking method. In order to clarify mechanisms of cross flow in such tight lattice rod bundles, the TPFIT is applied to simulate water-steam two-phase flow in two modeled subchannels. Attention is focused on instantaneous fluctuation characteristics of cross flow. With the calculation of correlation coefficients between differential pressure and gas/liquid mixing coefficients, time scales of cross flow are evaluated, and effects of mixing section length, flow pattern and gap spacing on correlation coefficients are investigated. Differences in mechanism between gas and liquid cross flows are pointed out.

Critical Power Characteristics in 37-rod Tight Lattice Bundles under Transient Conditions

Critical power characteristics in the postulated abnormal transient processes that may be possibly met in the operation of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) were investigated for the design of the FLWR core. Transient Boiling Transition (BT) tests were carried out using two sets of 37-rod tight lattice rod bundles (rod diameter: 13 mm; rod clearance: 1.3 mm or 1.0 mm) at Japan Atomic Energy Agency (JAEA) under the conditions covering the FLWR operating condition (Pex = 7:2 MPa, Tin = 556 K) for mass velocity G = 400-800 kg/(m2 s). For the postulated power increase and flow decrease transients, no obvious change of the critical power against the steady one was observed. The traditional quasi-steady characteristic was confirmed to be working for the postulated power increase and flow decrease transients. The experiments were analyzed with TRAC-BF1 code, where the JAEA newest critical power correlation for the tight lattice rod bundles was implemented for the BT judgment. The TRAC-BF1 code showed good prediction for the occurrence or the non occurrence of the BT and for the exact BT starting time. The traditional quasi-steady state prediction of the BT in transient process was confirmed to be applicable for the postulated abnormal transient processes in the tight lattice rod bundles.

ICONE15-10332 DEVELOPMENT OF DESIGN TECHNOLOGY ON THERMAL-HYDRAULIC PERFORMANCE IN TIGHT-LATTICE ROD BUNDLES : II- ROD BOWING EFFECT ON BOILING TRANSITION

A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we will describe the critical power characteristics in a 37-rod tight-lattice bundle with rod-bowing under both steady and transient states. It is observed that no matter it is run under a steady or a transient state, boiling transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle. Steady critical power increases monotonically with the increase of mass velocity, with the decrease of inlet water temperature and with the decrease of exit pressure. These trends are same as those in the base case test without rod-bowing. The steady critical power with rod-bowing is about 10% lower than that without rod-bowing. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with TRAC-BF1 code. The TRAC-BF1 code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time within the accuracy of critical power correlation. Traditional quasi - steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight lattice bundle with rod - bowing.

Application of the Post-Boiling-Transition Standard to Licensing Analysis

To ensure fuel integrity, light water reactor cores are designed to avoid the onset of boiling transition (BT) inside the fuel assembly that leads to a deterioration of the heat transfer characteristics and subsequent excessive rise of the fuel-cladding temperature in the anticipated operational occurrences (AOOs). However, some boiling water reactors’ AOO events result in immediate scram or suppression of the reactor power due to an increase in the reactor coolant void fraction. Recent studies show that a short duration of dryout inside the fuel assembly only leads to a small rise in the fuel-cladding temperature and thus does not pose a threat to fuel integrity. Many tests on BT and an improved comprehension of its mechanism have led to the development of a methodology to appropriately assess the fuel-cladding temperature after BT has been reached. The Standards Committee of the Atomic Energy Society of Japan has therefore proposed a cladding temperature criterion after BT. Applying the post-BT standard enables the value of the operating limit minimum critical power ratio (OLMCPR) to be decreased by allowing for a short duration of dryout. We calculated the fuel-cladding temperature and dryout duration in the load rejection condition without a bypass event. The calculated results show that both the fuel-cladding temperature and dryout duration meet the post-BT standard in the case of a small OLMCPR, which is determined by the loss of feedwater heating. This enables a more efficient reactor core to be designed by applying the post-BT standard to licensing analysis. The possibility of applying a post-BT standard is demonstrated from the results of this work.

OS8-04 稠密37本バンドル熱特性試験とギャップ幅効果の評価(OS8 軽水炉・新型炉・核燃料サイクル)

A thermal/hydraulic feasibility project for Reduced-Moderation Water Reactor (RMWR) was started since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes are being carried out. In this paper, results of the large-scale thermal-hydraulic tests which simulate the RMWR bundle are summarized. Especially, gap width effect on critical power which was investigated using two 37-rod bundle test sections with 1.3 and 1.0mm in gap width is reported. As a result, it is confirmed that (1) fundamental characteristics of flow parameter impact on critical power are similar between the results using two test sections with 1.3mm and 1.0mm in gap width, (2) effect of cross-sectional power distribution could be expressed by the power factor of subchannel where BT (Boiling Transition) would occur, (3) quasi-steady-state predictive method for BT is applicable for postulated transients and (4) proposed evaluation method of pressure drop could yield good predictions in the tight-lattice bundle.

Current Status of the Post Boiling Transition Research in Japan

Development of rewetting correlation formula was the key to predict fuel-cladding temperature after Boiling Transition (BT). Japanese BWR utilities and vendors performed some tests of rewetting and made two rewetting correlation formulas. The effect on fuel integrity after BT depends on temperature of fuel rod and time of dryout. Main cause of losing fuel integrity during BWR's Anticipated Operational Occurrences (AOO) after BT is embrittlement of the claddings due to oxidation. Ballooning of fuel rod is excepted because its pressure boundary isn't broken. In Japan, the Standards Committee of Atomic Energy Society of Japan (AESJ) is making post BT standard. This standard provides guidelines based on the latest knowledge to judge fuel integrity in case of BT and the validity of reusing the fuel assembly that experienced BT in BWRs.

Development of a Transient Boiling Transition Analysis Method Based on a Film Flow Model

A new single-channel, transient boiling transition (BT) prediction method based on a film flow model has been developed for a core thermal-hydraulic code. This method could predict onset and location of dryout and rewetting under transient conditions mechanically based on the dryout criterion and with consideration of the spacer effect. The developed method was applied to analysis of steady-state and transient BT experiments using BWR fuel bundle mockups for verification. Comparisons between calculated results and experimental data showed that the developed method tended to predict occurrence of rewetting earlier, however, onset time of BT and maximum rod surface temperature were well predicted within 0.6 s and 20°C, respectively. Moreover, it was confirmed that consideration of the spacer effect on liquid film flow rate on the rod surface was required to predict dryout phenomena accurately under transient conditions.

Theoretical Consideration of a Boiling Transition Model for the Low Flow Region

To introduce a theoretical model of boiling transition under low flow conditions, which is related closely to start-up condition in natural circulation BWR's, consideration on the nature of boiling transition models are carried out on the basis of previous experiments and theoretical studies. Two local models, the “Stability of dry patch” and the “Liquid Sublayer Dryout” model are investigated in detail according to their applicability and capability. The boiling transition model “Stability of dry patch” is based on the force balance. To enclose this model, parameters are assumed, which are discussed according to the system conditions of this study. The boiling transition model “Liquid Sublayer Dryout” is based on the energy balance and was extended to the annular flow region. The flow parameters of these models were determined by a drift flux model. The numerical prediction of boiling transition heat flux confirms with the experimental results under low flow condition using the liquid sublayer model and the dry patch model. However both models axe showing discrepancies in the prediction for larger mass fluxes, related to the inaccuracy in the prediction of the complex flow condition.

Theoretical Consideration of a Boiling Transition Model for the Low Flow Region.

To introduce a theoretical model of boiling transition under low flow conditions, which is related closely to start-up condition in natural circulation BWR's, consideration on the nature of boiling transition models are carried out on the basis of previous experiments and theoretical studies. Two local models, the “Stability of dry patch” and the “Liquid Sublayer Dryout” model are investigated in detail according to their applicability and capability. The boiling transition model “Stability of dry patch” is based on the force balance. To enclose this model, parameters are assumed, which are discussed according to the system conditions of this study. The boiling transition model “Liquid Sublayer Dryout” is based on the energy balance and was extended to the annular flow region. The flow parameters of these models were determined by a drift flux model. The numerical prediction of boiling transition heat flux confirms with the experimental results under low flow condition using the liquid sublayer model and the dry patch model. However both models axe showing discrepancies in the prediction for larger mass fluxes, related to the inaccuracy in the prediction of the complex flow condition.

Critical Power Experimental Analysis Using Subchannel Analysis Code. Unknown Parameter Estimation in Spacer Model.

In order to predict the critical power of a BWR rod bundle by calculation, we have developed a fuel rod spacer effect model for quantitative estimation of the spacer effect on film flow on a fuel rod surface. Unknown parameters introduced in the spacer effect model are determined, based on critical power test data conducted at the Toshiba 4×4 bundle test loop. In this parameter estimation procedure, a boiling transition (BT) rod position effect is included, using critical power test data in which BT is expected to occur at different rod positions. The spacer model developed in the present program has been implemented in the film flow model subchannel analysis code CRIP version 2, and the calculated critical power is compared with the 4×4 bundle test data. The results of measurement and analysis show good agreement.

An experimental study on rewetting phenomena in transient conditions of bwrs

It is known that rod temperature rise after boiling transition (BT) is not excursive and that the peak cladding temperature (PCT) is suppressed by rewetting to return to nucleate boiling, even if BT occurs under severe conditions exceeding abnormal operational transients for a BWR. The purpose of this study is to develop and verify the rewetting correlation. The rewetting correlation was developed based on single rod data, as a function of quality, mass flux, pressure and heat flux. The transient thermal-hydraulic code used in the BWR design analysis (SCAT) with this rewetting correlation was compared with transient rod temperature result after the occurence of BT obtianed by the 8×8 and 4×4 rod bundle. It is concluded that the transient code with the developed rewetting correlation predicts the PCT conservatively, and the rewetting time well.

Study on post-BT heat transfer in a full scale bwr ( [formula omitted]) rod bundle

Post-BT heat transfer experiments on a full scale BWR ( 8 × 8) rod bundle were conducted at the Nuclear Power Engineering Test Center (NUPEC), which is sponsored by the Ministry of International Trade and Industry (MITI). Demineralized water was used as a test fluid. The experimental conditions were: Mass flux: 284 to 1562 kg/m 2s, Pressure: 7.15 MPa, Inlet subcooling: 50 kJ/kg. Test results indicate that the rod temperature rise after boiling transition (BT) is not “excursive”. Comparisons with several existing post-BT correlations indicate that the Groeneveld, The Dougall-Rohsenow and the Condie-Bengston correlations tend to be conservative in the intermediate range before the onset of fully developed film boiling, while the Koizumi and the Sugawara correlations predict well over the complete Post-BT range.

Prediction of Transient Boiling Transition in BWR Fuel Rod Bundles by Subchannel Analysis Program MENUETT

Transient CHF (critical heat flux) tests of a 4 X 4 rod bundle were analyzed by the subchannel analysis program MENUETT. MENUETT is based on a non-equilibrium, five equation, two-phase flow model and is available both for steady state and transient analyses. Turbulent mixing and void drift effects are taken into account to calculate cross flows in fuel rod bundles. The tendency of calculated subchannel mass fluxes and qualities agreed with experimental data. By using a critical quality correlation obtained from steady state CHF data, the position of the earliest boiling transition could be predicted regardless of non-uniform axial heat flux distributions. This transition occurrence time was predicted within a difference of 0.1∼0.3 s from the experimental time. MENUETT applicability was confirmed for transient calculations predicting thermalhydraulic behavior in bundles.