[en] Prediction of dryout point is experimentally investigated with deionized water upflowing through narrow annular channel with 1.0 mm and 1.5 mm gap respectively. The annulus with narrow gap is bilaterally heated by AC current power supply. The experimental conditions covered a range of pressure from 0.8 to 3.5 MPa, mass flux of 26.6 to 68.8 kg·m_²·s^-1 and wall heat flux of 5 to 50 kW·m^-2. The location of dryout is obtained by observing a sudden rise in surface temperature. Kutateladze correlation is cited and modified to predict the location of dryout and proved to be not a proper one. Considering in detail the effects of geometry of annuli, pressure, mass flux and heat flux on dryout, an empirical correction is finally developed to predict dryout point in narrow annular gap under low flow condition, which has a good agreement with experimental data. (author)

[en] Based on the the droplet-diffusion model by Kirillov and Smogalev (1969, 1972), a new analytical model of dryout point prediction in the steam-water flow for bilaterally and uniformly heated narrow annular gap was developed. Comparison of the present model predictions with experimental results indicated that a good agreement in accuracy for the experimental parametric range (pressure from 0.8 to 3.5 MPa, mass flux of 60.39 to 135.6 kg·^-2·s^-1 and the heat flus of 50 kW·m^-2. Prediction of dryout point was experimentally investigated with deionized water upflowing through narrow annular channel with 1.0 mm and 1.5 mm gap heated by AC power supply. (author)

[en] Highlights: • A thermal-hydraulic analysis code named TSACO-HCCB TBM was developed. • The code was verified by comparing with RELAP5. • The design basis accident in-vessel LOCA analysis was performed with this code. • This article is useful for design and operation of helium cooling system. -- Abstract: In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system has been developed using Fortran 90 language, and the simulation has been performed for the cooling system of the Chinese helium cooled ceramic breeder test blanket module (CH HCCB TBM). The semi-implicit finite difference technique was adopted for the solution of the dynamic behavior of helium cooling system. Furthermore, a detailed illustration of the numerical solution for heat structures and critical model was presented. The code was verified by the comparison of RELAP5 code with the same initial condition, boundary condition, heat transfer and flow friction models. The TBM inlet/outlet temperatures and pressure drop were obtained and the results simulated by TSACO were shown in good agreement with those by RELAP5. Thereafter, the design basis accident in-vessel loss of coolant accident (LOCA), was investigated for the CH HCCB TBM cooling system. The critical flow model was also verified by comparing with RELAP5 code. The results indicated that the TBM can be cooled down effectively. The vacuum vessel (VV) pressure and the mass of helium spilled into the VV maintained below the design limits with a large margin

[en] Thermohydraulics characteristics in the secondary side of AP1000 steam generator (SG) are simulated based on the porous media models. The drift flux two-phase flow model coupled with a simplified flow boiling model is utilized. The heat transfer from primary side fluid to secondary side fluid is calculated three-dimensionally during iterations. The resistances caused by downcomer, tube bundle, support plates and primary separators are considered. Three-dimensional distributions of velocity, temperature, pressure, enthalpy, density, void fraction and flow vapor quality are obtained from the calculation by using the CFD code ANSYS FLUENT. Flow-induced vibration (FIV) damage is analyzed based on the cross flow velocity over the U-bend region of the outmost U-tube. The most severe FIV damages occur at the angles of −0.544 rad on the cold side and 0.353 rad on the hot side with maximum cross flow energies of 1145.2 J/m³ and 658.9 J/m³, respectively. Fouling is expected to deposit at the bottom of tube bundle since the velocity there is close to zero. The flow vapor qualities of mixture flowing into separators vary from each other significantly, with the maximum and minimum flow vapor quality in separators of 0.659 and 0.073, which is a severe challenge to the capacity design of separators. -- Highlights: • Secondary side of steam generator is simulated with porous media model. • Heat transfer from primary to secondary side is taken into account. • Localized flow characteristics of secondary side are obtained. • Parameters to analyze FIV damage, fouling and separator load are obtained

[en] Leak rate calculation is the foundation of Leak-Before-Break (LBB) technology application in Pressurized Water Reactor (PWR). In the paper, a leak rate Mathcad calculation code with different critical flow mathematical assumptions was completed. The code calculation results were contrasted from the published experimental data. The compared results show that the code calculation results are coincident with experimental data. However, the theoretical results are greater than experimental data with different crack L/D, stagnation pressures and subcooled temperatures in case of ignoring friction effect. While the crack friction effect is considered, the calculated results are well in accordance with the experimental data. Also, the different pressure drops are obtained and studied with variations of important parameters in detail. It demonstrates that the friction effect is a significant factor and must be considered in the crack leak rate calculation. The Mathcad code can be used to calculate the crack leak rate and provide application foundation of LBB in PWR pipe system. -- Highlights: • The leak rate mathematical models of LBB were studied and modified. • Mathcad programming of leak rate calculation and comparing with experimental data. • Studies of different initial conditions and friction effect. • Formed Mathcad code can be used to calculate the crack leak rate accurately

[en] In the present study, thermal-hydraulics characteristics of AP1000 passive residual heat removal heat exchanger (PRHR-HX) at initial operating stage were analyzed based on the porous media models. The data predicated by RELAP5 under the condition of the station blackout was employed as the inlet flow rate and temperature boundary of CFD calculation. The heat transfer from the primary side coolant to the in-containment refueling water storage tank (IRWST) side fluid was calculated in a three-dimensional geometry during iterations, and the distributed resistances were added into the C-type tube bundle regions. Three-dimensional distributions of velocity and temperature in the IRWST were calculated by the CFD code ANSYS FLUENT. The primary temperature, heat transfer coefficients of two sides and the heat transfer were obtained using the coupled heat transfer between the primary side and the IRWST side. The simulation results indicated that the water temperature rises gradually which leads to a thermal stratification phenomenon in the tank and the heat transfer capability decreases with an increase of water temperature. The present results indicated that the method containing coupled heat transfer from the primary side fluid to IRWST side fluid and porous media model is a suitable approach to study the transient thermal-hydraulics of PRHR/IRWST

[en] Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void worth increased with the MA weight fraction, while the Doppler feedback and the effective delayed neutron fraction decreased with the increase of the MA weight fraction

[en] Highlights: • Spraying characteristics in the pressurizer of Pressurized Water Reactor were tested. • Spray droplet collector was fabricated to obtain a 2-D spray flux distribution graph. • A photography chamber was specially implemented to measure the spray drop size. • An exponent was proposed for the discharge coefficient of the pressure-swirl nozzle. • A better empirical method for Probability Density Function of drop size was proposed. - Abstract: Spraying system in the pressurizer of Pressurized Water Reactor (PWR) power plant system is of great importance for system pressure control. An experimental study on the spray characteristics, including mass flow rate, spray flux distribution, spray cone angle and drop size spectrum, was conducted. A testing loop with nine swirling nozzles was established for the study. In order to measure the spray cone angle and drop size spectrum, two original devices including a spray droplet collector and a photographic chamber were designed and employed. The former was used to collect the spray droplet along the cross-section diameter, and the latter was made to isolate and measure the targeted spray droplet. Based on the experimental data, the curves of flow rate and spray cone angle versus nozzle pressure drop were obtained. Several typical spray flux distributions were derived and the results indicated that the flux distribution changes significantly with even small pressure changes. Thus, it was proposed that instability of the spray flux distribution should be considered in the pressurizer. Based on the spray drop pictures recorded by the high speed camera, Probability Density Function (PDF) of the drop size was obtained and compared with four ‘standard’ empirical distributions. It was found that the Nukiyama–Tanasawa distribution provides a better fit to the experimental PDF of the spray drop size. The present work introduces the experimental methodology and results of spray behaviour of the nozzle in pressurizer. The work is expected to be helpful for the optimization design of spraying systems

[en] Highlights: • An innovative passive system design concept was proposed. • Remote trigger control was adopted in the new passive system. • Transient characteristic of new PRHRS was studied during normal operation, SBO and FLB. • New design type is successful. - Abstract: An innovative passive safety system design concept for CPR 1000 nuclear power plant (NPP), proposed based on the original passive residual heat removal system (PRHRS) design, is presented and studied in this paper. The new type PRHRS avoids use of trigger valve inside the containment and adds an external water tank as the trip of PRHRS operation. The new design concept can realize functions of system remote control and trigger, while the external water tank can provide emergency water for steam generator (SG) in case of loss of feed water accident. Also, the advantages to use a valve and a tank located externally to containment are including the more reliability in implementing valve in pipes at low pressure, possibility operating manually the valve to start the system, refilling the tank in case of long time cooling and maintaining the valve easily. Best-estimate transient simulation code Relap5/MOD3.4 is applied to study the behavior of new design PRHRS and transient characteristics of primary loop system during normal condition and accident conditions. The transient processes of Station Black-Out (SBO) and Feed-water Line Break (FLB) accidents are studied to verify the function of new design PRHRS. Results show that the effect of valve position change from inside to outside containment can be ignored for the new design PRHRS in normal operation and the external valve can trip the operation of PRHRS in case of accidents. The new design PRHRS can remove core residual heat from primary loop effectively through establishing the stable natural circulations in the primary loop and PRHRS loop when the SBO and FLB accidents occur. It also can supply the emergency water to the SG in the event of FLB accident and realize safe reactor shutdown. Results indicate that the new design concept PRHRS in this work is potential and successful in future application

[en] Highlights: • The safety performance of the optimized HCSB blanket for CFETR has been investigated using RELAP5. • In-vessel LOCA and ex-vessel LOCA under ITER-like condition are investigated. • The parametric analyses are carried out. • The optimized HCSB blanket is designed with sufficient decay heat removal capability. - Abstract: A conceptual design of helium cooled solid breeder (HCSB) blanket, one of three blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR), has been performed recently. The transient analysis for different possible accidents should be carried out to assess its safety performance. In this paper, the ITER-like conditions are adopted since the associated system for CFETR is missing, such as helium cooling system, plasma shutdown condition, pump behavior, etc. The complete model is recreated including the optimized typical outboard HCSB blanket (NO.12) and its ancillary helium cooling loop, and accident analyses of two loss of coolant accident (LOCA) cases are investigated using RELAP5. The influences of different break areas under in-vessel LOCA are compared, and the accident consequence after small area break is chosen to be investigated. Regarding the ex-vessel LOCA, the influences of different break locations are thoroughly analyzed. Since the plasma cannot terminate passively after ex-vessel pipe break, the plasma termination behaviors are investigated with different shutdown time. The computational results show that with the safety criteria for ITER the HCSB blanket can be cooled down effectively by the helium cooling system (HCS) and the integrity of pressure barriers can be guaranteed for both accidents.